[Federal Register Volume 60, Number 128 (Wednesday, July 5, 1995)]
[Notices]
[Pages 35058-35091]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-16249]
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NUCLEAR REGULATORY COMMISSION
Applications and Amendments to Facility Operating Licenses
Involving No Significant Hazards Considerations; Biweekly Notice
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from June 10, 1995, through June 22, 1995. The
last biweekly notice was published on June 21, 1995 (60 FR 32359).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration.
[[Page 35059]]
Under the Commission's regulations in 10 CFR 50.92, this means that
operation of the facility in accordance with the proposed amendment
would not (1) involve a significant increase in the probability or
consequences of an accident previously evaluated; or (2) create the
possibility of a new or different kind of accident from any accident
previously evaluated; or (3) involve a significant reduction in a
margin of safety. The basis for this proposed determination for each
amendment request is shown below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and should cite the publication date and page
number of this Federal Register notice. Written comments may also be
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.
The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By August 4, 1995, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington DC, by the above
date. Where petitions are filed during the last 10 days of the notice
period, it is requested that the petitioner promptly so inform the
Commission by a toll-free telephone call to Western Union at 1-(800)
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator
should be given Datagram Identification Number N1023 and the
[[Page 35060]]
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and
to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos.
1, 2, and 3, Maricopa County, Arizona
Date of amendment requests: May 2, 1995.
Description of amendment requests: The proposed amendment would
remove from the technical specifications (TS) plant elevations for the
minimum water volume required in the spent fuel pool (SFP) and relocate
them to site procedures. This proposed TS amendment also includes two
changes to correct administrative errors in the TS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis about the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change eliminates the plant elevations from TS
Figure 3.1-1, ``Minimum Borated Water Volumes'' for the SFP. The
change is administrative in nature and does not involve any
modifications to plant equipment or affected plant operation. The
required volume of water in the SFP is identified on the figure and
will remain unchanged by this amendment. This request relocates the
plant elevations to site procedures where they will be controlled in
accordance with the provisions of 10 CFR 50.59.
The removal of the reference to Table 3.8-2 in the Unit 3 TS
3.8.4.1 and adding the word ``containment'' to the Unit [2] TS
4.6.3.1 are administrative change[s] and do not involve any
modifications to plant equipment or affect plant operation. These
administrative changes do not affect the scope or intent of any test
within the TS.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change eliminates the plant elevations from TS
Figure 3.1-1, ``Minimum Borated Water Volumes'' for the SFP. The
change is administrative in nature and does not involve any
modifications to plant equipment or affect plant operation. The
removal of plant elevations from the figure does not cause any
change in the method by which any safety-related system performs its
function. The required volume of water in the SFP is identified on
the figure and will remain unchanged by this amendment.
The removal of the reference to Table 3.8-2 in the Unit 3 TS
3.8.4.1 and adding the word ``containment'' to the Unit 2 TS 4.6.3.1
are administrative changes and do not involve any modifications to
plant equipment or affect plant operation. These administrative
changes do not affect the scope or intent of any test within the TS.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change eliminates the plant elevations from TS
Figure 3.1-1, ``Minimum Borated Water Volumes'' for the SFP. The
change is administrative in nature and does not involve any
modifications to plant equipment or affect plant operation. The
required volume of water in the SFP is identified on the figure and
will remain unchanged by this amendment.
The removal of the reference to Table 3.8-2 in the Unit 3 TS
3.8.4.1 and adding the word ``containment'' to the Unit 2 TS 4.6.3.1
are administrative changes and do not involve any modifications to
plant equipment or affect plant operation. These administrative
changes do not affect the scope or intent of any test within the TS.
Therefore, based upon the above, the proposed change does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004.
Attorney for licensees: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999.
NRC Project Director: William H. Bateman.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,
Maryland
Date of amendments request: June 2, 1995.
Description of amendments request: The proposed amendments would
revise the pressurizer safety valve setpoint tolerance ``as-found''
acceptance criterion to +2%/-1% for the valve with the lower setpoint
(RC-200) and plus or minus 2% for the valve with the upper setpoint
(RC-201). The ``as-left'' setpoint tolerance will remain plus or minus
1% for both valves.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The pressurizer safety valves are used to prevent exceeding the
Reactor Coolant System (RCS) pressure safety limit. The proposed
change to increase the pressurizer safety valve setpoint tolerance
for the ``as-found'' acceptance criteria from [plus or minus]1% to
+2%/-1% for the valve with the lower pressure setpoint, and [plus or
minus] 2% for the valve with the upper pressure setpoint, does not
affect any initiating event. The proposed change does not affect the
consequences of the previously evaluated design basis accidents as
the new safety valve setpoint tolerances are bounded by the
assumptions in the safety analysis. Therefore, the proposed change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Would not create the possibility of a new or different type
of accident from any accident previously evaluated.
The proposed change to increase the ``as-found'' setpoint
tolerances does not involve any changes in equipment or the function
of these safety valves. The proposed change does not represent a
change in the configuration or operation of the plant. The test
method for the pressurizer safety valves will remain the same. The
increase in the setpoint tolerances does not create any new accident
initiator. Therefore, the proposed change does not create the
possibility of a new or different type of accident from any accident
previously evaluated.
[[Page 35061]]
3. Would not involve a significant reduction in a margin of
safety.
The pressure safety limit for the RCS protects the structural
integrity of the system from failure due to overpressurization. The
pressurizer safety valves are used to prevent the RCS pressure from
exceeding the safety limit. The proposed change to the pressurizer
safety valve setpoint tolerances will continue to prevent the RCS
pressure from exceeding the design safety limit during any design
basis event. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Ledyard B. Marsh.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,
Maryland
Date of amendments request: June 6, 1995.
Description of amendments request: The proposed amendments would
revise the Calvert Cliffs Nuclear Plant Units 1 and 2 Technical
Specifications, extending certain 18-month frequency surveillances to a
refueling interval (nominally 24 months, not to exceed 30 months).
Systems and equipment affected are the Reactor Protective System (RPS),
Engineered Safety Features Actuation System (ESFAS), Power-Operated
Relief Valve (PORV) actuation instruments, Low Temperature Overpressure
Protection (LTOP)-related instruments, Remote Shutdown Panel
instruments, Post-Accident Monitoring (PAM) instruments, Containment
Sump Level instruments, and Radiation Monitoring instruments.
This amendment request would extend the nominal surveillance
interval requirement from 18 months to a refueling interval (nominally
24 months, not to exceed 30 months) for instrument channel
calibrations, RPS and ESFAS total bypass function operability
verification, RPS and ESFAS time response tests, ESFAS Manual Trip
Button channel functional tests, and ESFAS Automatic Actuation Logic
Channel Functional Tests. Calvert Cliffs has been operating on a 24-
month fuel cycle since July 1987 (Unit 2) and July 1988 (Unit 1),
performing some Technical Specification surveillances, such as the ones
described here, during mid-cycle outages. The request is the last of a
series of proposed license amendments that would eliminate the need for
planned mid-cycle outages to perform required surveillances.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The proposed change would extend surveillance intervals for
Reactor Protective System (RPS), Engineered Safety Features
Actuation System (ESFAS), Power-Operated Relief Valve (PORV), Low
Temperature Overpressure Protection (LTOP), Remote Shutdown, Post-
Accident Monitoring (PAM), Radiation Monitoring, and Containment
Sump Level Instruments.
The purpose of the RPS is to effect a rapid reactor shutdown if
any one or a combination of conditions deviates from a pre-selected
operating range. The system functions to protect the core and the
Reactor Coolant System pressure boundary. The purpose of the ESFAS
is to actuate equipment which protects the public and plant
personnel from the accidental release of radioactive fission
products if an accident occurs, including a loss-of-coolant
incident, main steam line break, or loss of feedwater incident. The
safety features function to localize, control mitigate, and
terminate such incidents in order to minimize radiation exposure to
the general public. The Post-Accident Monitoring instruments provide
the Control Room operators with primary information necessary to
take manual actions, as necessary, in response to design basis
events, and to verify proper system response to plant conditions and
operator actions. The purpose of the Remote Shutdown System is to
provide plant parameter indications to operators on a Remote
Shutdown Panel to be used while placing and maintaining the plant in
a safe shutdown condition in the event the Control Room is
uninhabitable. The indications are used to verify proper system
response to plant conditions and operator actions. The LTOP System
protects against Reactor Coolant System overpressurization at low
temperatures by a combination of administrative controls and
hardware. The hardware includes two Power-Operated Relief Valves
with variable pressurizer pressure setpoints when operating in the
LTOP operating parameter region. The Containment Sump High Level
Alarm System provides an alarm in the Control Room for a containment
sump to provide one of the available indications of excessive RCS
leakage during normal plant operation. The Containment Area High
Range Radiation Monitoring System provides an indication of high
radiation levels in containment. The Containment Purge System
actuates equipment to prevent the release of radioactive material to
the environment in the event of a reactor coolant leak, a shielding
failure, or a fuel pin failure when the reactor vessel head is
removed.
The instruments in each of the systems described above are
designed to be used in response to an accident. Failure of any of
these systems is not an initiator for any previously evaluated
accident. Therefore, the proposed change would not involve an
increase in the probability of an accident previously evaluated.
Many of the instruments addressed in this license amendment
request will have or have recently had a new brand of sensor
installed. The effect of the increased surveillance interval with
the new sensors was analyzed. The new sensors do not effect the
physical design description of the plant, any design or functional
requirements, or surveillances. The proposed Technical Specification
change extending the surveillance interval from 18 months to a
refueling interval (nominally 24 months, not to exceed 30 months)
does not physically change the plant, change any design or
functional requirements, or effect the surveillances themselves.
Analysis has shown that no trip setpoints need to be changed, and
operator indications will continue to be accurate for control of
plant parameters to effect a safe shutdown. The equipment will
continue to perform as designed to mitigate the consequences of
accidents. Therefore, the proposed change would not involve a
significant increase in the consequences of an accident. [* * *]
Therefore, the proposed change would not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Would not create the possibility of a new or different type
of accident from any accident previously evaluated.
The proposed change to increase the interval RPS, ESFAS, PORV,
LTOP, Remote Shutdown, PAM, Radiation Monitoring, and Containment
Sump Level instrument surveillances from 18 months to a refueling
interval (nominally 24 months, not to exceed 30 months) does not
involve a significant change in the design or operation of the
plant. No hardware is being added to the plant as part of the
proposed change. Some detector upgrades in specific plant systems to
enhance the performance of those systems have been or will be
performed. However, those upgrades were evaluated and deemed
acceptable under 10 CFR 50.59 and are not part of this request. The
Reactor Protective System, Engineered Safety Features Actuation
System, Power-Operated Relief Valve, Low Temperature Overpressure
Protection, Containment Sump Level, one Radiation Monitoring
actuation setpoints will not be changed. Analysis has shown that the
remote shutdown and PAM indications will continue to be accurate.
The proposed change will not introduce any new accident initiators.
Therefore, this change does not create the possibility of a new or
different type of accident from any previously evaluated.
3. Does operation of the facility in accordance with the
proposed amendment
[[Page 35062]]
involve a significant reduction in a margin of safety?
The impact of the surveillance interval extension request was
evaluated for each Technical Specification-related safety function
for each of the RPS, ESFAS, PORV, LTOP, Remote Shutdown, PAM,
Radiation Monitoring, and Containment Sump Level instruments
addressed by this submittal. In all cases, parameters specified in
the related accident analysis were determined to be unaffected by
the surveillance interval extension, and no accident analyses limits
required changes. The Reactor Protective System, Engineered Safety
Features Actuation System, Power-Operated Relief Valve, Low
Temperature Overpressure Protection, Containment Sump Level, and
Radiation Monitoring actuation setpoints will not be changed.
Analysis has shown that the remote shutdown and PAM indications will
continue to be accurate. The methods for detection of degraded
instrument operation have not been changed, and remote shutdown and
PAM operator indications will continue to provide adequate accuracy.
The methods for detection of degraded instrument operation have not
been changed, and remote shutdown and PAM operator indications will
continue to provide adequate accuracy.
The proposed change does not affect the operation of the systems
involved. The surveillance interval extension will not affect the
design of the systems, and methods for detection of degraded
instrument operation will continue to identify operation problems
between calibrations. Therefore, the proposed change does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Ledyard B. Marsh.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,
Maryland
Date of amendments request: June 9, 1995.
Description of amendments request: The proposed amendments revise
the Calvert Cliffs Nuclear Power Plant Radiological Effluent Technical
Specifications (RETS) consistent with Generic Letter (GL),
``Implementation of Programmatic Controls For Radiological Effluent
Technical Specifications in the Administrative Controls Section of the
Technical Specifications and the Relocation of Procedural Details of
RETS to the Offsite Dose Calculation Manual or the Process Control
Program (Generic Letter 89-01),'' dated January 31, 1989, and the
Improved Standard Technical Specifications for Combustion Engineering
Plants published in NUREG-1432, as modified by Mr. W. T. Russell's
letter of October 25, 1993, ``Content of Standard Technical
Specifications,'' to the Improved Technical Specification Owners Group
Chairpersons. Changes for relocating the procedural details of the
current RETS to the Offsite Dose Control Manual (ODCM) has been
prepared in accordance with the proposed changes to the Administrative
Controls section of the Technical Specifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed change has been evaluated against the standards in
10 CFR 50.92 and has been determined to not involve a significant
hazards consideration, in that operation of the facility in
accordance with the proposed amendments:
1. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The proposed changes will provide human factor improvements for
the Technical Specifications by relocating existing procedural
details of the current Radiological Effluent Technical
Specifications to the Offsite Dose Control Manual (ODCM). Procedural
details for solid radioactive wastes will be relocated to the
Process Control Program. The proposed amendment (1) incorporates
programmatic controls in the Administrative Controls section of the
Technical Specifications that satisfy the requirements of 10 CFR
20.1302, 40 CFR Part 190, 10 CFR 50.36a, 10 CFR Part 50, Appendix I,
and our current Technical Specifications; (2) relocates the existing
procedural details in current specifications involving radioactive
effluent monitoring instrumentation, the control of liquid and
gaseous effluents, equipment requirements for liquid and gaseous
effluents, radiological environmental monitoring, and radiological
reporting details from the Technical Specifications to the ODCM; (3)
simplifies the associated reporting requirements; (4) simplifies the
administrative controls for changes to the ODCM; and (5) updates the
definitions of the ODCM consistent with these changes.
Relocating existing requirements and eliminating requirements
which duplicate regulatory requirements provide Technical
Specifications which are easier to use. Because existing
requirements are relocated to established programs where changes to
those programs are controlled by regulatory requirements, there is
no reduction in commitment and adequate control is still maintained.
Likewise, the elimination of requirements which duplicate regulatory
requirements enhances the usability of the Technical Specifications
without reducing commitments. The additional improvements being
proposed neither add nor delete requirements, but merely clarify and
improve the readability and understanding of the Technical
Specifications. Since the requirements remain the same, these
changes only affect the method of presentation, and as such, would
not affect possible initiating events for accidents previously
evaluated or any system functional requirement.
Furthermore, no safety-related equipment, safety function, or
plant operation will be altered as a result of this proposed change.
The changes are unrelated to the initiation and mitigation of
accidents and equipment malfunctions addressed in the Updated Final
Safety Analysis Report.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Would not create the possibility of a new or different type
of accident from any accident previously evaluated.
Transferring the procedural details of radiological effluent
monitoring and reporting from the Technical Specifications to the
ODCM has no impact on plant operation or safety. No safety-related
equipment, safety function, or plant operation will be altered as a
result of this proposed change. No changes to plant components or
structures are introduced which could create new accidents or
malfunctions not previously evaluated.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Would not involve a significant reduction in a margin of
safety.
The margin of safety associated with the affected Technical
Specifications is to provide assurance that the releases of
radioactive materials during actual or potential releases of liquid
or gaseous effluents do not exceed the limits of 10 CFR Part 20.
This license amendment request relocates the methodology and
parameters used to ensure that the 10 CFR Part 20 limits are
maintained, but does not change any of these requirements. Thus, no
methodology and parameters for controlling radioactive effluent
releases will be changed.
The procedural details of the current Radiological Effluent
Technical Specifications will be transferred to the ODCM and
replaced with programmatic controls consistent with regulatory
requirements, including controls on revisions to the ODCM. Thus, no
requirements or controls will be reduced.
The proposed revisions to the reporting requirements for
Radiological Effluent Release Report and the revision from the old
10 CFR 20.106 requirements to the new 10 CFR 20.1302 have no impact
on plant systems, plant operations or accident precursors. The
changes to the effluent
[[Page 35063]]
reporting requirements and the updated reference to 10 CFR 20.1302 do
not change either the means of controlling radioactive releases or
the effluent release limits. Therefore, there will be no change in
the types and amounts of effluents that will be released, nor will
there be an increase in individual or cumulative radiation exposures
to any member of the public.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Ledyard B. Marsh.
Carolina Power & Light Company, Docket No. 50-261, H.B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: June 3, 1995.
Description of amendment request: The requested Technical
Specification (TS) change clarifies the definition of operability of
the charging pumps by adding a footnote to TS Section 3.2.2.a that
states that the connectibility of the emergency power sources is not
required for charging pump operability.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
This change request does not involve a significant hazards
consideration for the following reasons.
1. The requested change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The requested change clarifies that the emergency power
sources are not required for the operability of the charging pumps.
Operation of the charging pumps is not considered in the assumptions
for initiation of any analyzed accident and is not credited for
accident mitigation in any analyzed accidents in the safety analysis
report. Therefore, the availability of emergency power sources to
the charging pumps does not affect the probability of occurrence or
consequences of an analyzed accident in the safety analysis report.
2. The requested change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated. The requested change clarifies that the emergency power
sources are not required for the operability of the charging pumps.
The design requirements of the charging pumps to provide reactor
coolant inventory and boron inventory control are not changed. The
operability of the emergency power source to the charging pumps is
not a precursor to any accident scenario. Failure of the charging
pumps is bounded by the plant design which strips the charging pumps
from the emergency buses under certain conditions. Since the change
does not involve changes in the operation of the plant, or physical
or equipment changes or involve controls for accident mitigation
equipment, the requested change will not create the possibility of
new or different kind of accident from any accident previously
evaluated.
3. The requested change clarifies that the emergency power
sources are not required for the operability of the charging pumps.
Since the charging pumps are stripped from the emergency buses in
the event of a loss of power and safety injection, emergency power
sources to the charging pumps are not guaranteed to mitigate the
consequences of an analyzed accident. As a result, no credit is
taken for the charging function in analyzed accidents and the margin
of safety as described in the safety analysis report is unchanged.
Therefore, the requested change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550.
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602.
NRC Project Director: David B. Matthews.
Commonwealth Edison Company, Docket Nos. 50-454 and 50-455, Byron
Station, Unit Nos. 1 and 2, Ogle County, Illinois
Docket Nos. 50-456 and 50-457, Braidwood Station, Unit Nos. 1 and 2,
Will County, Illinois
Date of amendment request: February 21, 1995.
Description of amendment request: The proposed amendments would
revise Byron and Braidwood technical specifications associated with the
reactor coolant system (RCS) resistance temperature detectors (RTDs)
used to obtain hot and cold leg temperatures. The amendments are
required because of proposed modification that will remove the existing
RTDs and their associated piping and valves and replace them with dual
element fast response RTDs mounted in the thermowells welded directly
in the RCS loop piping.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed modification replaces the existing bypass piping
system with thermowell-mounted RTDs. Because the hot leg RTDs are
mounted directly in the scoops, temperature measurement inaccuracies
caused by imbalances in the flow scoop sample flow are eliminated.
The method of measuring coolant temperature with thermowell-mounted
fast response RTDs has been analyzed to be at least as effective as
the RTD bypass system. With the thermowells welded into the existing
RCS hot and cold leg nozzles and the elimination of the bypass
piping, the number of pressure boundary welds has been significantly
reduced, resulting in a reduced probability of a small break LOCA
[Loss of Coolant Accident].
The RTD response time is incorporated in the safety analyses. In
particular, RTD response time is modeled in the OT[DELTA]T [Over
Temperature Delta Temperature] and OP[DELTA]T [Over Pressure Delta
Temperature] trip functions. The overall response time modeled in
the safety analyses for the existing RTD bypass piping system is 8
seconds. The overall response time is the elapsed time from the time
the temperature change in the RCS exceeds the trip setpoint until
the rods are free to fall. More specifically, 6 seconds is modeled
as a first order lag term and 2 seconds as pure delay on the reactor
trip signal. The 6 second lag term includes such factors as: RTD
bypass piping fluid transport delay, RTD bypass piping thermal lag,
RTD response time, and RTD electronic filtering. The 2 second delay
on reactor trip addresses such factors as electronics delay, trip
breakers and gripper release.
Signal conditioning (filtering) of the individual loop [DELTA]T
and Tavg signals is represented by [time constants utilized in
the lag compensator for DELTA T] and [time constant utilized in the
measured Tavg lag compensator], respectively, in the OT[DELTA]T
and OP[DELTA]T equations in Technical Specification Table 2.2-1.
With the current bypass manifold system, the filter is not required
since the existing RTDs do not respond rapidly to local temperature
variances within the reactor coolant loop. The bypass piping and
manifold provide adequate mixing of the coolant, eliminating any
local temperature variances. Therefore, the values of [time
constants utilized in the lag compensator for DELTA T] and [time
[[Page 35064]]
constant utilized in the measured Tavg lag compensator] are
currently specified as 0 seconds, effectively turning off the
electronic filter. The new fast response RTDs may respond to
temperature spikes which are not representative of actual RCS bulk
fluid temperature. Signal conditioning may be required to eliminate
these temperature spikes. Although, the current Technical
Specifications do not provide for any signal conditioning, the 8
second total response time used in safety analyses has sufficient
margin to account for a typical 2 second time constant for signal
conditioning. Industry experience has shown that a 2 second filter
is adequate in eliminating the spikes.
The proposed fast response RTD/thermowell system also has an
overall response time of 8 seconds. However, the time distribution
for the parameters differ between the existing and proposed designs.
The existing design includes a transport time for RCS fluid to reach
the RTD, located in the manifold. The RTDs are directly immersed
into the coolant, providing a fast response. The new design no
longer has the transport delay. However, because the RTDs are
mounted in thermowells, the response time of the RTD/thermowell
combination will be increased over the existing system.
The effects of a redistribution of the time responses between
the total lag term (pipe transport delay, RTD response and
electronic filter delay) and electronics delay term have been
evaluated. Westinghouse completed a Safety Evaluation SECL-95-015,
``OT[DELTA]T and OP[DELTA]T Reactor Trip Response Time Safety
Evaluation'' to support the revision to the time requirements. The
evaluation concludes that, as long as the total response time
remains [less than or equal to] 8 seconds, the safety analyses
acceptance criteria continue to be met. The OT[DELTA]T and
OP[DELTA]T trip functions are unaffected by the change.
The following Updated Final Safety Analysis Report (UFSAR)
Chapter 15 events trip on OT[DELTA]T: Loss of Electric Load/Turbine
Trip, Uncontrolled RCCA Bank Withdrawal at Power, CVCS Malfunction
that Results in a Decrease in the Boron Concentration in the Reactor
Coolant, and Inadvertent Opening of a Pressurizer Safety or Relief
Valve. In addition, the following events trip on OP[DELTA]T:
Steamline Break at Hot Full Power for Core Response, and Steamline
Break Superheat Analysis. These events have been reviewed for a
change in the distribution of time responses for OT[DELTA]T and
OP[DELTA]T. The review concludes that the time response
redistribution did not result in a minimum DNBR lower than the
safety analyses limit, did not result in a fuel centerline melt, nor
did the superheated steam releases change from those currently
existing. Therefore, the radiological consequences for these events
do not increase as a result of the less restrictive time response
breakdown. Thus, the proposed amendment does not result in an
increase in the probability or consequences of a previously
evaluated accident.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The OT[DELTA]T and OP[DELTA]T trip functions are unaffected by
the change. Electronic filtering of the RTD signal has been
included, changing the dynamic compensation term of OT[DELTA]T and
OP[DELTA]T setpoint equations. No other changes to the setpoint
equation result from the proposed modification.
The added 7300 hardware is compatible with the existing 7300
electronic hardware now used. All changes to the 7300 protection
cabinets have been qualified. The proposed system is functionally
equivalent to the existing one. The proposed modification has been
reviewed for conformance with the Institute of Electrical and
Electronics Engineers (IEEE) 279-1971 criteria, associated General
Design Criteria, Regulatory Guides, and other applicable industry
standards. The single failure criterion is satisfied by the proposed
modification, since the independence of redundant protection sets is
maintained. The new RTD/thermowell system meets the equipment
seismic and environmental qualification requirements of IEEE
standards 344-1975 and 323-1974, respectively. The proposed changes
do not affect the protection system capabilities to initiate a
reactor trip. The 2 of 4 voting coincidence logic of the protection
sets is maintained. Therefore, the proposed modification meets all
appropriate IEEE criteria, industry standards and other guidelines.
In addition, the RTD outputs are used for rod control, turbine
runback, pressurizer level and other control systems. These control
systems receive the signal after it has been processed at the 7300
cabinets and are therefore unaffected by the proposed modification.
The design and installation of the thermowells is in accordance
with the American Society of Mechanical Engineers (ASME) Code
requirements. However, should a thermowell fail at the RCS pressure
boundary, the resulting accident is enveloped by current design
basis accident analyses. Thus, implementation of the proposed
amendment does not create the possibility of a new or different kind
of accident from any of those previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The 7300 protection cabinets calculate individual loop [DELTA]T
and Tavg, based on the output of the RTDs. These values are
used in the OT[DELTA]T and OP[DELTA]T reactor protection trip
signals. Electronic filtering of the RTD signal will be included,
changing the dynamic compensation term of OT[DELTA]T and OP[DELTA]T
setpoint equations. No other changes to the setpoint equation result
from the proposed modification. Although the total response time
used as input into the safety analyses is unaffected by the proposed
modification, the distribution of response times between the total
lag (pipe transport delay, RTD response and electronic filter delay)
and the electronic delay has changed. The UFSAR events which rely on
OT[DELTA]T and OP[DELTA]T trips have been evaluated. The evaluation
concludes that the safety analyses acceptance criteria continue to
be met, since the total response time is consistent with the safety
analyses. The OT[DELTA]T and OP[DELTA]T trips function in the same
manner to terminate DNB-related transients. The reliability of the
reactor protection system is unaffected by this change. Thus, the
proposed modification does not involve a significant reduction in
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Robert A. Capra.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1
and 2, Will County, Illinois
Date of amendment request: May 17, 1995.
Description of amendment request: The proposed amendment would
modify the technical specifications to allow steam generator tubes to
be repaired using the tungsten inert gas (TIG) welded sleeve process
developed by ABB Combustion Engineering (ABB/CE), remove the ability to
repair steam generator tubes using the Babcock & Wilcox Nuclear
Technologies (BWNT) kinetically welded sleeve process, and increase the
requirement to inspect the number of sleeved tubes from 3 percent of
the total number of sleeved tubes in all four steam generators (SGs) or
all sleeved tubes in one steam generator to 20 percent of each sleeve
design installed. The proposed amendments would also delete the
requirement to conduct additional corrosion testing to establish the
design life for the BWNT kinetically welded sleeve in the presence of a
crevice.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or
[[Page 35065]]
consequences of an accident previously evaluated.
The proposed amendment allows the ABB/CE TIG welded tubesheet
sleeves and tube support plate sleeves to be used as an alternate
tube repair method for Byron and Braidwood Units 1 and 2 Steam
Generators (SGs). The sleeve configuration was designed and analyzed
in accordance with the criteria of Regulatory Guide (RG) 1.121 and
Section III of the ASME Code. Fatigue and stress analyses of the
sleeved tube assemblies produce acceptable results for both types of
sleeves as documented in ABB/CE Licensing Report CEN-621-P, Revision
00, ``Commonwealth Edison Byron and Braidwood Unit 1 & 2 Steam
Generator Tube Repair Using Leak Tight Sleeves, FINAL REPORT,''
April 1995. Mechanical testing has shown that the structural
strength of the sleeves under normal, faulted, and upset conditions
is within the acceptable limits specified in RG 1.121. Leakage rate
testing for the tube sleeves has demonstrated that primary to
secondary leakage is not expected during any plant condition. The
consequences of leakage through the sleeved region of the tube is
fully bounded by the existing steam generator tube rupture (SGTR)
analysis included in the Byron and Braidwood Updated Final Safety
Analysis Report (UFSAR).
The current Technical Specification 3.4.6.2.c primary to
secondary leakage limit of 150 gallons per day (gpd) through any one
SG ensures that SG tube integrity is maintained in the event of main
steam line break (MSLB) or loss of coolant accident (LOCA). The RG
1.121 criteria for establishing operational leakage rate limits
require a plant shutdown based upon a leak-before-break
consideration to detect a free span crack before a potential tube
rupture. The 150 gpd limit will continue to allow for early leakage
detection and require a plant shutdown in the event of the
occurrence of an unexpected crack resulting in leakage that exceeds
the TS limit.
The sleeves are designed to allow inservice inspection of the
pressure retaining portions of the sleeve and parent tube. Inservice
inspection is performed on all sleeves following installation to
ensure that each sleeve has been properly installed and is
structurally sound. Periodic inspections are performed in subsequent
refuel outages to monitor sleeve degradation on a sample basis. The
eddy current technique used for inspection will be capable of
detecting both axial and circumferential flaws. A 20% sample of the
sleeves are inspected each refuel outage. In the event that an
imperfection exceeding the repair limit is detected an additional
20% sample will be inspected. The inspection scope is expanded to
100% of the sleeves should a repairable defect be found in the
second sample. Tubes that contain defects in a sleeve, which exceed
the repair limit, will be removed from service. This ensures that
sleeve and tube structural integrity is maintained.
The proposed TS change to support the installation of TIG welded
sleeves does not adversely impact any previously evaluated design
basis accident. The effect of sleeve installation on the performance
of the SG was analyzed for heat transfer, flow restriction, and
steam generation capacity. The sleeves reduce the risk of primary to
secondary leakage in the SG. The installation of ABB/CE sleeve
results in a hydraulic flow restriction that is dependent on the
number and types of sleeves installed. The reduction in primary
system flow rate is a small percentage of the flow rate reduction
seen from plugging one tube and is a preferable alternative when
considering core margins based on minimum reactor coolant system
flow rates. The sleeving installation will result in a resistance to
primary coolant flow through the tube for other evaluated accidents.
The results of the analyses and testing, as well as industry
operating experience, demonstrate that the sleeve assembly is an
acceptable means of maintaining tube integrity. In summary,
installation of sleeves does not substantially affect the primary
system flow rate or the heat transfer capability of the steam
generators.
The sleeve sample size has been increased from 3% of the sleeved
tubes in all four steam generators to include an eddy current
inspection of a minimum of 20% of each sleeve design installed.
Increasing the sample size of the sleeves to be inspected will
increase the monitoring of tubes using sleeves for any further
degradation while they remain in service. If the sample identifies a
sleeve with an imperfection of greater than the repair limit, an
additional 20% of the sleeves shall be inspected. The sleeves that
have identified imperfections of greater than the repair limit shall
be removed from service. Increasing the monitoring of the sleeves
will assist in the early detection of a tube or sleeve imperfection
and limit the probability of occurrence of an accident previously
evaluated in the UFSAR.
Installation of the sleeves can be used to repair degraded tubes
by returning the condition of the tubes to their original design
basis condition for tube integrity and leak tightness during all
plant conditions. The tube bundle overall structural and leakage
integrity will be increased with the installation of the sleeves
reducing the risk of primary to secondary leakage in the SG while
maintaining acceptable reactor coolant system flow rates. Therefore
sleeving will not increase the probability of occurrence of an
accident previously evaluated.
Removal of the BWNT kinetically welded sleeve process as an
approved SG tube repair methodology and not completing the
additional corrosion testing necessary to establish the design life
for the BWNT kinetically welded sleeve in the presence of a crevice
will have no affect on plant operations. There are currently no BWNT
kinetically welded sleeves installed in the Byron or Braidwood SGs.
Had there been, plant operations would have still been bounded by
the existing SGTR analysis in the Byron and Braidwood UFSAR.
Therefore, these proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The implementation of the proposed sleeving process will not
introduce significant or adverse changes to the plant design basis.
Stress and fatigue analyses of the repair has shown the ASME Code
and RG 1.121 allowable values are met. Implementation of TIG welded
sleeving maintains overall tube bundle structural and leakage
integrity at a level consistent with that of the originally supplied
tubing. Leak and mechanical testing of the sleeves support the
conclusions that the sleeve retains both structural and leakage
integrity during all conditions. Repair of a tube with a sleeve does
not provide a mechanism that result in an accident outside of the
area affected by the sleeve.
Any hypothetical accident as a result of potential tube or
sleeve degradation in the repaired portion of the tube is bounded by
the existing SGTR analysis. The SGTR analysis accounts for the
installation of sleeves and the impact on current plugging level
analyses. The sleeve design does not affect any other component or
location of the tube outside of the immediate area repaired.
The current Technical Specification 3.4.6.2.c primary to
secondary leakage limit of 150 gpd through any one SG ensures that
SG tube integrity is maintained in the event of an MSLB or LOCA. The
limit will provide for leakage detection and a plant shutdown in the
event of the occurrence of an unexpected single crack resulting in
excessive tube leakage. The leakage limit also provides for early
detection and a plant shutdown prior to a postulated crack reaching
critical crack lengths for MSLB conditions.
Inservice inspections are performed following sleeve
installation to ensure proper weld fusion has occurred to maintain
structural integrity. The post installation inspection also serves
as baseline data to be used for comparison during future
inspections. Periodic eddy current inspections monitor the pressure
retaining portions of the sleeve and parent tube for degradation.
Eddy current techniques will be employed that are sensitive to axial
and circumferential degradation.
Increasing the sample size of tubes repaired using either
sleeving process during each scheduled inservice inspection will
increase the monitoring of these tubes for any further degradation.
The improved monitoring and evaluation of the tube and the sleeves
assures tube structural integrity is maintained or the tube is
removed for service.
Corrosion testing of typical sleeve-tube configurations was
performed to evaluate local stresses, sleeve life, and resistance to
primary and secondary side corrosion. The tests were performed on
stress relieved and as-welded (non-stress relieved) sleeve-tube
joints. Using the corrosion test data in conjunction with finite
element analyses of the local stress, the stress relieved joint life
was determined to be in excess of 40 years. The ABB/CE TIG welded
sleeve operating experience in the industry has shown no sleeve
failures due to service induced degradation in sleeves that were
installed with acceptable inspection results. This experience
includes the stress relieved and
[[Page 35066]]
as-welded sleeve configurations. ComEd will stress relieve all sleeves
at Byron and Braidwood as specified in the Technical Report.
Removal of the BWNT kinetically welded sleeve process as an
approved SG tube repair methodology and not completing the
additional corrosion testing necessary to establish the design life
for the BWNT kinetically welded sleeve in the presence of a crevice
will not create the possibility of a new or different type of
accident from any accident previously evaluated. Repair of an SG
tube with a BWNT kinetically welded sleeve would not have provided a
mechanism that resulted in an accident outside of the area affected
by the sleeve. Any hypothetical accident as a result of potential
tube or sleeve degradation in the repaired portion of the tube would
have been bounded by the existing SGTR analysis. The SGTR analysis
accounts for the installation of sleeves and the impact on current
plugging level analyses. The sleeve design does not affect any other
component or location of the tube outside of the immediate area
repaired. Furthermore, there are currently no BWNT kinetically
welded sleeves installed in the Byron or Braidwood SGs.
Therefore, the proposed changes do not create the possibility of
a new or different type of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The TIG welded sleeving repair of degraded steam generator tubes
has been shown by analysis to restore the integrity of the tube
bundle to its original design basis condition. The safety factors
used in the design of the sleeves for the repair of degraded tubes
are consistent with the safety factors in the ASME Boiler and
Pressure Vessel Code used in steam generator design. The design of
the ABB/CE SG sleeves has been verified by testing to preclude
leakage during normal and postulated accident conditions.
The portions of the installed sleeve assembly which represents
the reactor coolant pressure boundary can be monitored for the
initiation and progression of sleeve/tube wall degradation, thus
satisfying the requirement of RG 1.83. The portion of the SG tube
bridged by the sleeve joints is effectively removed from the
pressure boundary, and the sleeve then forms the new pressure
boundary. The sleeve enhances the safety of the plant by
reestablishing the protective boundaries of the steam generator.
Keeping the tube in service with the use of a sleeve instead of
plugging the tube and removing it from service increases the heat
transfer efficiency of the steam generator. During each scheduled
inservice inspection, each sleeve inspected and found to have
unacceptable degradation shall be removed from service. The effect
on the design transients and the accident analyses have been
reviewed based on the installation of sleeves equal to the tube
plugging level coincident with the minimum reactor coolant flow
rate. Evaluation of the installation of sleeves was based on the
determination that LOCA evaluations for the licensed minimum reactor
coolant flow bound the combined effect of tube plugging and sleeving
up to an equivalent of the actual plugging limit. Sleeving results
in a fractional amount of the plugging limitation of one tube and is
a preferable alternative when considering core margins based on
minimum reactor coolant system flow rates. The sleeving installation
will result in a resistance to primary coolant flow through the
tube. The primary coolant flow through the ruptured tube is reduced
by the influence of the installed sleeve, thereby reducing the
consequences to the public due to a SGTR event.
A SG sleeve removes an indication of a possible leak source from
the reactor coolant system (RCS) pressure boundary, eliminating the
potential of a primary-to-secondary leak. The structural integrity
of the tube is maintained by the sleeve and sleeve-to-tube joint.
Installation of either tube sheet or tube support plate sleeves
will increase the protective boundaries of the steam generators and
will not reduce the margin of safety.
Removal of the BWNT kinetically welded sleeve process as an
approved SG tube repair methodology and not completing the
additional corrosion testing necessary to establish the design life
for the BWNT kinetically welded sleeve in the presence of a crevice
will not result in a reduction in the margin of safety. There are
currently no BWNT kinetically welded sleeves installed in the Byron
or Braidwood SGs. SG tube integrity will be maintained by applying
an alternate NRC approved repair methodology or removing the SG tube
from service by plugging.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Robert A. Capra.
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: April 11, 1995.
Description of amendment request: The proposed amendments would
allow a one-time extension of specific LaSalle, Units 1 and 2, 18 month
Technical Specification Surveillance Requirements to allow surveillance
testing to coincide with the LaSalle, Unit 1, seventh refueling outage
(L1R07). The shutdown for L1R07 has been rescheduled from September
1995 until early 1996. The proposed extensions apply to: Calibrations
and functional testing of isolation actuation instrumentation,
emergency core cooling system actuation instrumentation, and
recirculation pump trip actuation instrumentation; leakage testing of
reactor coolant system isolation valves; inspection of fire rated
seals; functional testing of mechanical snubbers; inspections of
emergency diesel generators; and testing of batteries, battery
chargers, and other electrical components.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated because:
The proposed change is temporary and allows a one-time extension
of specific surveillance requirements for Unit 1 Cycle 7 to allow
surveillance testing to coincide with the seventh refueling outage.
The proposed surveillance interval extension is short and will not
cause a significant reduction in system reliability nor affect the
ability of the systems to perform their design function. Current
monitoring of plant conditions and continuation of the surveillance
testing required during normal plant operation will continue to be
performed to ensure conformance with Technical Specification
operability requirements. Therefore, this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated because:
Extending the surveillance interval for the performance of
specific testing will not create the possibility of any new or
different kind of accidents. No changes are required to any system
configurations, plant equipment, or analyses. Therefore, this change
will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
(3) Involve a significant reduction in the margin of safety
because:
Surveillance interval extensions will not impact any plant
safety analyses since the assumptions used will remain unchanged.
The safety limits assumed in the accident analyses and the design
function of the equipment required to mitigate the consequences of
any postulated accidents will not be changed since only the
surveillance test interval is being extended. Historical performance
generally indicates a high degree of reliability, and surveillance
[[Page 35067]]
testing performed during normal plant operation will continue to be
performed to verify continued Operability of affected systems,
structures and components. Therefore, the plant will be maintained
within the analyzed limits, and the proposed extension will not
significantly reduce the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Robert A. Capra.
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: May 19, 1995.
Description of amendment request: The proposed amendments would
revise the technical specification requirement to verify each fire
protection valve is in the correct position at least once per 31 days.
The proposed change will retain a monthly visual inspection of the fire
protection valves that are accessible during plant operation. However,
the interval for visual surveillance of those valves considered not
accessible during plant operation will be changed to at least once per
18 months.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated because: The
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated in
the UFSAR [Updated Final Safety Analysis Report]. The proposed
change only changes the testing frequency for valves that are
inaccessible during power operation. A check of the LaSalle LER
database for the entire operating lifetime of LaSalle Units 1 and 2
was performed, and there has not been any instances in which any
Technical Specification related Fire Protection valves have been
found out of position. Therefore, the change to the frequency of
testing will have no affect on the capability of fire suppression
water systems, since all Technical Specification fire protection
valves, both accessible and inaccessible at power operation, have a
plant history of 100% correct valve lineup during monthly
surveillances. Additionally, all fire protection valves that are in
the fire suppression water flow path are either locked or seal wired
in the required position at all times. The change does not impact
the probability of any fire or other accident occurrence. Therefore,
the proposed change does not cause an increase in the probability or
consequences of an accident previously evaluated.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated because:
The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated
in the UFSAR. The proposed change only changes the testing frequency
for valves that are inaccessible during power operation. The change
to the frequency of testing will have no effect on the capability of
fire suppression water systems, since the valves, both accessible
and inaccessible at power operation, have a plant lifetime history
of 100% correct valve lineup during monthly surveillances.
Additionally, these valves are locked or sealed in the required
position at all times. The change does not alter the performance of
the fire suppression water system, and therefore introduces no new
failure modes. With no alteration or degradation to equipment or
system operation, the change introduces no new accident or
malfunction.
(3) Involve a significant reduction in the margin of safety
because:
The proposed change does not reduce the margin as defined in the
bases for any Technical Specification. The proposed change only
changes the testing frequency for all Technical Specification fire
protection valves that are inaccessible during power operation. The
plant history of 100% correct valve lineup for the Technical
Specification fire protection valves, combined with the fact that
these valves are always locked or sealed in the required position
ensures that the bases' minimum OPERABILITY requirements of the fire
suppression systems are met.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Robert A. Capra.
Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station, Units 1 and 2, Lake County, Illinois
Date of amendment request: May 31, 1995.
Description of amendment request: The proposed amendments would
revise the Technical Specifications and incorporate new acceptance
criteria for steam generator tubes with degradation in the tubesheet
roll expansion region.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability of occurrence or consequences of an accident
previously evaluated.
Application of the F* criteria to degraded steam generator tubes
will not affect any of the initiators or precursors of any accident
previously evaluated. Application of the proposed change will not
increase the likelihood that a transient initiating event will occur
because transients are initiated by equipment malfunction and/or
catastrophic system failure. The proposed change will allow a new
criteria to be applied to disposition steam generator tubes that are
degraded in the tubesheet roll transition region. The F* criteria
specify a minimum length of tubing which must be free from any
indication of degradation. Below the F* length, any type or size of
indication, including complete circumferential through wall
cracking, will not impact the structural integrity of the tube with
respect to pull out forces during normal operation or accident
conditions, and does not significantly affect the leakage behavior
of the tube. While the Zion UFSAR does not specifically address the
Feedwater Line Break (FLB) accident, the FLB event was used as the
limiting event in the evaluation of the F* criteria. The FLB
pressure differential of 2650 psi maximizes the axial loading on the
tube for pull out considerations and is bounding. In addition, the
close proximity of the tubesheet to the tube will prevent tube
rupture or collapse of the tube in the tubesheet span. Because
application of the F* criteria will ensure that degraded tubes will
provide the same structural integrity as an original undegraded tube
during normal operation and accident and accident conditions, the
probability of occurrence of an accident previously evaluated is not
significantly increased.
Application of the F* criteria will not significantly increase
the consequences of any accident previously evaluated. The F*
criteria ensure that sufficient length of undegraded tube exists to
maintain structural integrity and preclude significant leakage. Due
to the proximity of the tubesheet to the tube, any leakage from
degradations below the F* length would be negligible and would be
well below the Technical Specification limits established for steam
generator
[[Page 35068]]
leakage. Tube rupture as a result of indications below the F* distance
is precluded because the tubesheet prevents outward expansion of the
tube in response to internal pressure.
The relationship between the tubesheet region leak rate at the
most limiting postulated accident conditions relative to that for
normal plant operating conditions has been assessed. For the
postulated leak source within the roll expansion, increasing the
differential pressure on the tube on the tube wall increases the
driving head for the leak; however, it also increases the tube to
tubesheet loading.
For a leak source below the F* Distance, the maximum assumed
pressure differential results in an insignificant leak rate relative
to that which could be associated with normal plant operation. This
is a result of the increased tube to tubesheet loading associated
with the increased differential pressure. Thus for a circumferential
indication within the roll expansion that is left in service in
accordance with F* criteria, any leakage under accident conditions
would be less than that experienced under normal operating
conditions. Therefore, any leakage under accident conditions would
be less than the existing Technical Specification leakage limit,
which is consistent with accident analysis assumptions. Steam
generator tube integrity must be maintained under the postulated
loss of coolant accident condition of secondary-to-primary
differential pressure. Based on tube collapse strength
characteristics, the constraint provided to the tube by the
tubesheet gives a margin between the tube collapse strength and the
limiting secondary-to-primary differential pressure condition, even
in the presence of circumferential or axial indications. The maximum
secondary to primary differential pressure during a postulated LOCA
is 1005 psi. This value is significantly below the residual preload
between the tubes and the tube sheet. Therefore, no significant
secondary to primary leakage would be expected to occur.
In addition, the proposed changes will not affect the ability to
safely shut down the operating unit and mitigate the consequences of
an accident because the proposed changes will not necessitate
changes to the emergency procedures governing accident conditions or
plant recovery.
Administrative and typographical changes are proposed to correct
previous grammatical errors, to eliminate a parenthetical note that
could cause confusion when applying the proposed requirements, and
to make the terminology used in the Bases section consistent with
the definitions provided in Specification 4.3.1. Those proposed
changes will not increase the probability of occurrence or
consequence of any accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes to the Technical Specifications do not
involve the addition of any new or different types of safety related
equipment nor do they involve the operation of any equipment
required for safe operation of the facility in a manner different
from those addressed in the UFSAR. No safety related equipment or
function will be altered as a result of the proposed changes. Also,
the procedures governing normal plant operation and recovery from an
accident are not changed by the application of the F* criteria. The
F* criteria will allow the use of an alternate method to plugging or
sleeving to repair steam generator tubes with degradation in the
tubesheet region. The F* criteria ensure that both the structural
integrity and leak tight nature of the steam generator tube will be
equivalent to the original tube. Since no new failure modes or
mechanisms are introduced by the proposed changes, no new or
different type of accident is created.
Administrative and typographical changes are proposed to correct
previous grammatical errors, to eliminate a parenthetical note that
could cause confusion when applying the proposed requirements, and
to make the terminology used in the Bases section consistent with
the definitions provided in Specification 4.3.1. Those proposed
changes will not create the possibility of a new or different kind
of accident from those previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
Plant safety margins are established through Limiting Conditions
for Operation (LCOs), limiting safety system settings, and safety
limits specified in Technical Specifications. There will be no
changes to the LCOs, limiting safety system settings, or the safety
limits as a result of the proposed changes. Application of the F*
criteria will allow degraded steam generator tubes to be repaired by
an alternative method to plugging or sleeving. Steam generator tube
plugging decreases the total primary reactor coolant flow rate and
heat transfer capability of the steam generator. While steam
generator tube sleeving only slightly reduces the reactor coolant
flow rate, large numbers of sleeves can have a measurable effect on
flow rate and can complicate steam generator tube inspection
activities.
Application of the F* criteria will allow a repair method that
will restore the integrity of degraded steam generator tubes and
will not adversely affect primary system flow rate or heat transfer
capability. Application of the F* criteria will preserve the heat
transfer capability of the steam generators and will maintain the
design margins assumed in the analyses contained in the UFSAR. The
alternate repair method will also be less complicated, faster, and
will reduce personnel occupational exposure significantly. Based on
the above discussion it is concluded that the proposed changes will
not significantly reduce a margin of safety.
Administrative and typographical changes are proposed to correct
previous grammatical errors, to eliminate a parenthetical note that
could cause confusion when applying the proposed requirements, and
to make the terminology used in the Bases section consistent with
the definitions provided in Specification 4.3.1. Those proposed
changes will not impact any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Robert A. Capra.
Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas
Nuclear One, Unit Nos. 1 and 2 (ANO-1&2), Pope County, Arkansas
Date of amendment request: April 4, 1995.
Description of amendment request: The proposed amendments revise
requirements associated with the ventilation system that services both
the Unit 1 and Unit 2 control rooms.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1--Does Not Involve a Significant Increase in the Probability
or Consequences of an Accident Previously Evaluated.
The control room emergency ventilation and air conditioning
systems are not initiators of an accident previously evaluated.
Extension of the allowable outage time for one inoperable control
room emergency air conditioning system from 7 days to 30 days is
acceptable based on the low probability of an event occurring that
would require control room isolation and a concurrent or subsequent
failure of the remaining operable control room emergency air
conditioning system. An evaluation using probabilistic safety
assessment techniques has shown the frequency of this event to be at
an acceptably low level (4.67E-6/yr). The ANO-1 surveillance
requirements for the control room emergency ventilation and air
conditioning system has been updated for consistency with the ANO-2
requirements and are consistent with RG 1.52, March 1978, Revision
2. The relaxation in the ANO-2 Mode of Applicability for the control
room radiation monitoring instrumentation is acceptable based on the
fuel handling accident analysis dose consequences. The analysis
assumes that the control room emergency ventilation system is
actuated during a fuel handling accident in the containment
building. This analysis also shows that the dose consequences to the
control room operators are acceptable in the event of a fuel
handling analysis in the
[[Page 35069]]
auxiliary building, assuming that the normal control room ventilation
system only is in operation. When the unit is in Mode 5 or Mode 6
(with no handling of irradiated fuel in the containment building),
no accident condition has been identified that would require the
control room emergency ventilation system to actuate due to high
radiation. The remainder of the changes have been made for
consistency between the ANO-1 and ANO-2 TS and are considered to be
administrative in nature.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
Criterion 2--Does Not Create the Possibility of a New or Different Kind
of Accident from any Previously Evaluated
The control room emergency ventilation and air conditioning
systems are not accident initiators. The proposed changes introduce
no new mode of plant operation and no new possibility for an
accident is introduced by modifying the ANO-1 surveillance testing
requirements for the control room emergency ventilation and air
conditioning systems.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--Does Not Involve a Significant Reduction in the Margin of
Safety
With the exception of the AOT extension and the relaxation of
the ANO-2 Mode of Applicability for the control room radiation
monitoring instrumentation, all the ANO-1 and ANO-2 changes are
considered administrative or more restrictive and are intended to
clarify and make consistent the requirements of the control room
emergency habitability equipment. Although the AOT extension does
involve an incremental reduction in the margin of safety due to a
slight increase in the frequency of an event requiring control room
isolation, followed by failure of the operable emergency control
room chiller, a probabilistic safety assessment has shown this
slight increase in frequency (approximately 3.58E-6/yr) to be
acceptably low. The relaxation in the ANO-2 Mode of Applicability
for the control room radiation monitoring instrumentation is
acceptable based on the fuel handling accident analysis dose
consequences. The analysis assumes that the control room emergency
ventilation system is actuated during a fuel handling accident in
the containment building. This analysis also shows that the dose
consequences to the control room operators are acceptable in the
event of a fuel handling analysis [sic., accident] in the auxiliary
building, assuming that the normal control room ventilation system
only is in operation. When the unit is in Mode 5 or Mode 6 (with no
handling of irradiated fuel in the containment building), no
accident condition has been identified that would require the
control room emergency ventilation system to actuate due to high
radiation.
Therefore, this change does not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.
NRC Project Director: William D. Beckner.
Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas
Nuclear One, Unit Nos. 1 and 2 (ANO-1&2), Pope County, Arkansas
Date of amendment request: April 4, 1995.
Description of amendment request: The proposed amendments delete
requirements to perform inservice inspections of reactor coolant pump
flywheels at both Unit 1 and Unit 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1--Does Not Involve a Significant Increase in the Probability
or Consequences of an Accident Previously Evaluated.
Missile generation from a reactor coolant pump (RCP) flywheel
could damage the reactor coolant system, the containment, or other
equipment or systems important to safety. The fracture mechanics
analyses conducted to support the change shows that a preexisting
crack sized just below detection level will not grow to the flaw
size necessary to create flywheel missiles within the life of the
plant. This analysis conservatively assumes minimum material
properties, maximum flywheel accident speed, location of the flaw in
the highest stress area and a number of startup/shutdown cycles
eight times greater than expected. Since an existing flaw in the
flywheel will not grow to the allowable flaw size under normal
operating conditions or to the critical flaw size under LOCA
conditions over the life of the plant, elimination of inservice
inspections for such cracks during the plant's life will not involve
a significant increase in the probability of an accident previously
considered.
The proposed changes do not increase the amount of radioactive
material available for release or modify any systems used for
mitigation of such releases during accident conditions. Therefore,
these changes do not involve a significant increase in the
consequences of any accident previously evaluated.
Criterion 2--Does Not Create the Possibility of a New or Different Kind
of Accident from any Previously Evaluated
The proposed changes will not change the design, configuration,
or method of operation of the plant and therefore, will not create
the possibility of a new or different kind of accident from any
previously evaluated.
Criterion 3--Does Not Involve a Significant Reduction in the Margin of
Safety
Significant conservatisms have been used for calculating the
allowable flaw size, critical flaw size and crack growth rate in the
RCP flywheels. These include minimum material properties, maximum
flywheel accident speed, location of the flaw in the highest stress
area and a number of startup/shutdown cycles eight times greater
than expected. Since an existing flaw in the flywheel will not grow
to the allowable flaw size under normal operating conditions or to
the critical flaw size under LOCA conditions over the life of the
plant, elimination of inservice inspections for such cracks during
the plant's life will not involve a significant reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.
NRC Project Director: William D. Beckner.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: April 4, 1995.
Description of amendment request: The proposed amendment revises
surveillance requirements associated with the main turbine steam
valves.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1--Does Not Involve a Significant Increase in the Probability
or Consequences of an Accident Previously Evaluated.
Modifying the surveillance frequency of the main turbine-
generator (MTG) overspeed protection system introduces no new
failure mechanism for the machine, so the consequences, of a
postulated MTG overspeed event are no different than those
previously evaluated.
[[Page 35070]]
As explained in NUREG-1366, ``Improvements to Technical
Specifications Surveillance Requirements,'' the present surveillance
test frequency requirements were developed for fossil units and
carried over to nuclear units due to the similarity in design.
However, the particulate concentration, phosphate chemistry and
higher steam temperatures present in earlier fossil secondary
systems, which were major contributing factors to problems
identified by these tests, are not present in the Arkansas Nuclear
One-Unit 2 (ANO-2) secondary systems. The operating history of
turbine valves at ANO-2 is very good, with no failures identified
during the performance of overspeed protection system surveillance
testing. Therefore, that change does not involve a significant
increase in the probability of any accident previously evaluated.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Criterion 2--Does Not Create the Possibility of a New or Different Kind
of Accident from any Previously Evaluated.
Because the proposed changes do not alter the design,
configuration, or method of operation of the plant, they do not
create the possibility of a new or different kind of accident from
any previously evaluated.
Criterion 3--Does Not Involve a Significant Reduction in the Margin of
Safety.
These proposed changes do not alter the acceptance of any
surveillance requirements, alter any assumptions used in accident
analysis, change any actuation setpoints, nor allow operations in
any configuration not previously evaluated. This change in
surveillance frequency is based on an operating history of the
turbine overspeed protection system which indicates that reducing
the test frequency will have no adverse impact on the continued safe
operation of the unit.
Therefore, this change does not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: William D. Beckner.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida
Date of amendment request: May 31, 1995.
Description of amendment request: The proposed amendment would
revise the the Technical Specifications (TS) for the Crystal River Unit
3 to facilitate a 24 month operating cycle by changing the surveillance
interval for appropriate TS surveillance requirements that are
generally performed during a refueling outage. Additionally, the
functional description and the ``Allowable Value'' for three Reactor
Protection System and one Emergency Feedwater Initiation and Control
System setpoints would be revised. The quantitative limits for
determining the operational status of the reactor coolant pumps, the
main feedwater pumps, and the main turbine would be relocated from the
TS to the Final Safety Analysis Report (FSAR). The surveillance
associated with the high radiation setpoint for control room isolation
would also be changed to reflect that the setpoint is an ``approximate
value'' instead of an ``Allowable value''. The current specified
surveillance interval for some equipment and systems which were not re-
evaluated or which could not be justified by the evaluation process
would not be changed.
Specifically:
1. TS Surveillance Requirements (SR) 3.3.1.6, SR 3.3.5.3, SR
3.3.6.1, SR 3.3.9.2, SR 3.3.10.2, SR 3.3.11.3, SR 3.3.17.2, SR
3.3.18.2, and SR 3.9.2.2 would be revised to extend the surveillance
frequency from 18 to 24 months. Also, in TS SR 3.3.17.2 a note would be
added indicating the frequency for Function 12 is 18 months.
2. In TS Table 3.3.1-1,
(a) the Function for ``Reactor Coolant Pump Power Monitor (RCPPM)''
would be changed to ``Reactor Coolant Pumps,'' and the ``Allowable
Value'' column for this function would be revised to delete the
quantitative value and to indicate ``More than one pump tripped'',
(b) the Function for ``Main Turbine Trip (Control Oil Pressure)''
would be changed to ``Main Turbine,'' and the Allowable Value is
changed to ``Turbine Tripped'' and
(c) the Function for ``Loss of Both Main Feedwater Pumps (Control
Oil Pressure)'' would be changed to ``Main Feedwater Pumps,'' and the
Allowable Value is changed to ``Both Pumps Tripped''
3. In TS Table 3.3.11-1, Function 1.a would be changed from ``EFW
Initiation--Loss of MFW Pumps (Control Oil Pressure)'' to ``EFW
Initiation--Main Feedwater Pumps,'' and the Allowable Value is changed
to ``Both Pumps Tripped.''
4. In TS SR 3.3.16.3, the CHANNEL CALIBRATION setpoint would be
changed from an allowable value to an approximate setpoint.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence or the consequences of an accident
previously evaluated. The proposed amendment extends the interval
between successive refueling outage based surveillances to once
every 24 months for those surveillances evaluated herein and,
maintains the existing surveillance interval restriction for those
systems and equipment not evaluated for extension. The reliability
of systems and components relied upon to prevent or mitigate the
consequences of accidents previously evaluated is not degraded
beyond that obtained from the currently defined refueling outage
interval. Assurance of system and equipment availability is
maintained. This change does not involve any change to system or
equipment configuration. Therefore, this change does not increase
the probability of occurrence or the consequences of an accident
previously evaluated.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated. The
proposed amendment extends the interval between successive refueling
outage based surveillances to once every 24 months for those
surveillances evaluated herein and maintains the existing
surveillance interval restriction for those systems and equipment
not evaluated for extension. This change does not involve any change
to system or equipment configuration. Therefore, this change is
unrelated to the possibility of creating a new or different kind of
accident from any previously evaluated.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety. The proposed amendment extends the interval between
successive refueling outage based surveillances to once every 24
months for the surveillances evaluated herein, and maintains the
existing surveillance interval restriction for those systems and
equipment not evaluated for extension. The reliability of systems
and components is not degraded beyond that obtained from the
currently defined refueling outage interval. Assurance of system and
equipment availability is maintained.
Therefore, it is concluded that operation of the facility in
accordance with the proposed amendment does not involve a
significant reduction in a margin of safety. The proposed extension
of the refueling outage interval surveillances to once every 24
months does not degrade the reliability of systems and components
beyond that obtained from the currently defined refueling outage
interval.
[[Page 35071]]
Reliable performance of the systems and equipment effected by this
change has been demonstrated.
Implementation of the proposed amendment will maintain the
required level of assurance of system and equipment availability.
The surveillance interval for systems and equipment that have not
been evaluated for extension are excluded from this request. Thus,
operation of the facility in accordance with the proposed amendment
involves no significant hazards considerations.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 32629.
Attorney for licensee: A.H. Stephens, General Counsel, Florida
Power Corporation, MAC-A5D, P. O. Box 14042, St. Petersburg, Florida
33733.
NRC Project Director: David B. Matthews.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida.
Date of amendment request: May 31, 1995.
Description of amendment request: The proposed amendment would
revise the technical specifications (TS) for the Crystal River Nuclear
Plant Unit 3 (CR3) relating to the Once Through Steam Generator's
(OTSG's) tube inspection acceptance criteria. Currently, the TS specify
repair limit for removing steam generator tubes from service based on a
structural evaluation of a simplified model of tubes with uniform
through wall (T/W) thinning. A recent tube-pull examination at CR3
identified a number of low signal-to-noise (S/N) tube eddy current
indications. The licensee indicated that these S/N indications are a
substantially different morphology from the model used to develop the
current TS inspection and acceptance limit. As a result of the small
signal amplitude associated with these S/N indications, they cannot be
accurately sized by conventional bobbin coil phase angle. Therefore,
the licensee proposed an alternate methodology for dispositioning the
S/N indications. The proposed criteria would address both wear and
Inter-Granular-Attack (IGA) degradation mechanisms. Crack-like eddy
current indications are not included within the proposed scope.
Specifically, the licensee proposed to:
A. Revise TS 5.6.2.10.2, page 5.0-14, ``The results of each sample
inspection shall be classified into one of the following three
categories:'' to read: ``The results of each bobbin coil sample
inspection shall be classified into one of the following three
categories:''
B. Revise the Note in TS 5.6.2.10.2, page 5.0-14, ``In all
inspections, previously degraded tubes whose degradation has not been
spanned by a sleeve must exhibit a significant increase in the
applicable imperfection size measurement (> +0.5V bobbin coil amplitude
increase for S/N indications or >10% further wall penetration for all
other imperfections) to be included in the below percentage
calculations.''
C. Revise the sentence in TS 5.6.2.10.4.a.2, page 5.0-16, ``Eddy-
current* * *as imperfections'' to read: S/N indications with a bobbin
coil amplitude < 0.9V are considered imperfections. Other eddy current
testing indications below 20% of the nominal tube wall thickness, if
detectable, may also be considered as imperfections.
D. Revise TS 5.6.2.10.4.a.4, page 5.0-16, to read:
``Degraded Tube means a tube containing a S/N indication with a
bobbin coil amplitude 0.9V or other imperfection
20% of the nominal wall thickness caused by degradation
except where all such degradation has been spanned by the installation
of a sleeve.''
E. Add TS 5.6.2.10.4.a.7 ``Signal-to-Noise (S/N) indication means
an indication whose associated bobbin coil amplitude is < 5 times the
background noise, excluding indications located in the tube sheet
regions or indications determined to be other than a volumetric
morphology.''
F. Renumber 5.6.2.10.4.a.7 to 5.6.2.10.4.a.8, and revise to read:
Plugging/Sleeving Limit means the imperfection depth at or beyond which
the tube shall be restored to serviceability by the installation of a
sleeve or removed from service because it may become unserviceable
prior to the next inspection. The Limit for S/N indications is equal to
a bobbin coil amplitude of 2.5V, an axial extent of 0.33 inches, or a
circumferential extent of 0.6 inches. The Limit is equal to 40% of the
nominal tube or sleeve wall thickness for other imperfections. No more
than 5000 sleeves may be installed in each OTSG.
G. Renumber 5.6.2.10.4.a.8, and 9 to 5.6.2.10.4.a.9 and 10.
H. Revise TS 5.7.2.c.2, page 5.0-29, to read:
Location, bobbin coil amplitude, and axial and circumferential
extent (if determined) for each S/N indication and the location and
percent of wall thickness penetration for each other indication of an
imperfection, and
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change will not significantly increase the
probability or consequences of an accident previously evaluated. The
relevant accidents are excessive leakage or steam generator tube
rupture (as a consequence of MSLB [Main steam Line Break] or
otherwise).
RG [Regulatory Guide] 1.121 establishes a standard method for
demonstrating structural integrity under worse-than-DBE [design
basis Event] conditions. The existing TS is based on this RG. The S/
N disposition strategy continues to rely on this guidance. Current
TW sizing techniques would allow defects greater than the current TS
limit of 40% to remain in service since these techniques do not
accurately measure percent wall penetration for small volume
indications. The proposed disposition strategy is based in
measurable eddy current parameters of voltage, axial extent, and
circumferential extent shown to provide a higher confidence that
unacceptable flaws are removed from service. Therefore, the
probability of a Steam Generator Tube Rupture (SGTR) is not
increased and may well be decreased by implementation of this S/N
disposition strategy.
The probability of OTSG tube leakage during normal operation or
accident conditions is not adversely affected by the proposed S/N
disposition strategy. Operating history indicates essentially no
primary to secondary leakage through the OTSG tubes at CR-3. Growth
rate studies imply this trend could be expected to continue.
Therefore, current leakage limits are retained. Small volume
indications which might leak during worse-case FWLB [Feedwater Line
Break] conditions are addressed in the RG 1.121 evaluation. The
disposition strategy ensure these indications are removed from
service as part of the inservice inspection. Once detected, the
proposed criteria is at least as effective in determining those
indications which should be removed from service as are the existing
TS limits.
The S/N disposition strategy is an integral part of an overall
effort to better address these and similar phenomena in OTSGs.
2. The proposed change will not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The key `new or different' accidents addressed in this and
similar proposals is the potential for MSLB-induced multiple SGTR or
excessive primary-to-secondary leakage during such events. While
these events are addressed in CR-3 Emergency Operating Procedures
(EOPs), they are beyond those licensed for the facility.
However, as noted above, the probability of MSLB induced
multiple SGTR is reduced by more effective screening and plugging/
[[Page 35072]]
sleeving criteria. The probability of detection and identification of
tubes which should be removed from service is maintained or improved
by the S/N disposition strategy. The likelihood of adverse effects
from plugging sound tubes is reduced. The operation of the OTSG or
related structures, systems or components is otherwise unaffected.
3. The proposed change will not involve a significant reduction
to any margin of safety.
The margins of safety defined in RG 1.121, including the
required pressure used in the structural analysis, are retained. The
probability of detecting degradation is unchanged since bobbin coil
methods will continue to be the primary means of initial detection.
The probability of leakage remains acceptably small. The proposed S/
N disposition strategy is an enhancement to the inservice inspection
of OTSG tubing that will provide a higher level of confidence that
tubes exceeding the allowable limits are repaired while sound tubes
are left in service. Based upon results of the various growth rate
studies, the probability of an accident at the end of cycle is
essentially the same as the beginning.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 32629.
Attorney for licensee: A. H. Stephens, General Counsel, Florida
Power Corporation, MAC-A5D, P. O. Box 14042, St. Petersburg, Florida
33733.
NRC Project Director: David B. Matthews.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant Units 3 and 4, Dade County, Florida
Date of amendment request: June 19, 1995.
Description of amendment request: The licensee proposes to change
Turkey Point Units 3 and 4 Technical Specifications (TS) by separation
of the 24-hour emergency diesel generator (EDG) run and hot restart EDG
test from the loss-of-offsite-power load acceptance test. The licensee
revised the original amendment request dated March 30, 1995, by letters
dated May 5, 1995, and June 19, 1995.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which was previously presented in the Federal Register
(60 FR 27339, May 23, 1995). The licensee concluded that the proposed
license amendments' revisions do not alter the original conclusion that
no significant hazards considerations exist pursuant to 10 CFR 50.92.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request and its revisions involve no significant hazards
consideration.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
Attorney for licensee: J.R. Newman, Esquire, Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Project Director: David B. Matthews.
Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric
Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and
50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County,
Georgia
Date of amendment request: January 13, 1995, as supplemented by
letters dated April 5 and June 20, 1995.
Description of amendment request: The proposed amendments would
change the Facility Operating Licenses and their corresponding
Appendices A which contain the Technical Specifications (TS) to permit
the implementation of the power uprate program at the Edwin I. Hatch
Nuclear Plant, Units 1 and 2. The Hatch units are currently licensed
for operation at 2436 megawatts thermal (MWt). The proposed changes
would redefine the rated thermal power to 2558 MWt, which represents an
increase of 5% over the current licensed level in accordance with the
generic boiling water reactor (BWR) power uprate program established by
the General Electric Company (GE) and approved by the U.S. Nuclear
Regulatory Commission (NRC) staff in a letter from W. T. Russell, NRC,
to P. W. Marriott, GE, dated September 30, 1991. Implementation of the
proposed power uprate at Plant Hatch will result in an increase of
steam flow to approximately 106% of the current value but will require
no changes to the basic fuel design. Implementation of this proposed
power uprate will require minor modifications, such as resetting the
safety relief setpoints, as well as the calibration of plant
instrumentation to reflect the uprated power. Plant operating,
emergency, and other procedure changes will be made where necessary to
support uprated operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Will the changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
A. Rated Thermal Power is increased to 2558 MWt on page 3 of the
Unit 1 Operating License, page 4 of the Unit 2 Operating License,
and in Section 1.1 (Definitions) of the Units 1 and 2 Technical
Specifications.
Evaluation
The changes in the Operating Licenses and Technical
Specifications were evaluated and it was determined that the
probability (frequency of occurrence) of design basis accidents
occurring is not affected by the increased power level, as the
regulatory criteria established for plant equipment (e.g., ASME
Code, IEEE standards, NEMA standards, Regulatory Guide criteria)
will still be complied with at the uprated power level. Scram
setpoints (equipment settings that initiate automatic plant
shutdowns) will be established such that there is no significant
increase in scram frequency due to uprate. No new challenges to
safety-related equipment will result from power uprate.
The changes in consequences of hypothetical accidents which
would occur from 102% of the uprated power, compared to those
previously evaluated, are in all cases insignificant, because the
power uprate accident evaluations will not result in exceeding any
NRC-approved acceptance limits. Enclosure 4 of Reference 1, General
Electric Report NEDC-32405P, ``Power Uprate Safety Analysis for
Edwin I. Hatch Plant Units 1 and 2,'' December 1994, investigated
the spectrum of hypothetical accidents and transients, and showed
the plant's current regulatory criteria are satisfied at power
uprate. For example, in the area of core design, the fuel operating
limits will still be met at the uprated power level, and fuel reload
analyses will show plant transients meet the criteria accepted by
the NRC as specified in NEDO-24011, ``GESTAR II.'' Challenges to
fuel or emergency core cooling system (ECCS) performance were
evaluated (Section 4.2 of NEDC-32405P) and shown to still meet the
criteria of 10 [CFR] 50.46 and Appendix K. Challenges to the
containment were evaluated (Section 4.1 of NEDC-32405P) and shown to
still meet 10 CFR 50 Appendix A, Criterion 38, Long Term Cooling,
and Criterion 50, Containment. Radiological release events were
evaluated (Section 9.2 of NEDC-32405P) and shown to meet the
criteria of 10 CFR 100 (Unit 1 FSAR Chapter 14 and Unit 2 FSAR
Chapter 15).
The results of the analyses discussed above demonstrate that
operation at the power uprate level does not significantly increase
the probability or consequences of an accident previously evaluated.
B. The surveillance test discharge pressure for the standby
liquid control pump at 41.2 gpm is increased from 1190 psig to 1201
psig. This value appears in Surveillance Requirement (SR) 3.1.7.7
and the
[[Page 35073]]
corresponding Bases Section B 3.1.7 in the Unit 1 and Unit 2 Technical
Specifications.
Evaluation
Power uprate operation will result in a 30 psi increase in
reactor operating pressure. As will be discussed in these proposed
changes, several pressure-dependent setpoints (including safety
relief valve [SRV] setpoints) will be increased to preserve current
margins. Increasing the pressure 11 psi, at which a 41.2 gpm flow
rate is developed, assures continued conformance to anticipated
transient without scram (ATWS) criteria at uprated conditions. The
surveillance test pressure is based on the maximum pressure for an
ATWS event during the time period when the standby liquid control
pump is in operation. Section 6.5 of NEDC-32405P discusses the
capability of these positive displacement pumps. A small increase in
the SRV setpoints will have no effect on the rated injection flow to
the reactor. This change, therefore, will not increase the
probability or consequences of a previously evaluated accident.
C. The reactor vessel steam dome high pressure allowable value
for reactor protection system (RPS) instrumentation is increased 31
psi, consistent with the nominal pressure increase for power uprate.
The allowable value appears in Section 3.3.1.1, Table 3.3.1.1-1,
Function 3, in the Unit 1 and Unit 2 Technical Specifications.
Evaluation
The reactor vessel steam dome high pressure scram limit is
increased because the steam dome operating pressure is increased.
Operating pressure for uprated power is increased to assure that
satisfactory reactor pressure control is maintained. The operating
pressure was chosen on the basis of steam line pressure drop
characteristics and the steam flow capability of the turbine.
Satisfactory reactor pressure control requires an adequate flow
margin between the uprated operating condition and the steam flow
capability of the turbine control valves at their maximum stroke. An
operating dome pressure of 1035 psig, which is 30 psi higher than
the current operating dome pressure, is expected. Therefore, the
high pressure scram is increased approximately the same amount to
preserve existing margins to reactor trips.
The high pressure scram terminates a pressurization transient
not terminated by direct scram or high neutron flux scram. The
setting is maintained above the nominal reactor vessel operating
pressure and below the specified analytical trip limit used in the
safety analyses. The revised high pressure scram setpoint will
preserve the hierarchy of pressure setpoints. This means that the
high pressure scram setpoint will remain below the opening setpoint
of the SRVs. The SRV nominal setpoints are also increased 30 psi, as
discussed in Item G below. This hierarchy of setpoints provides
assurance that the probability of opening more than one SRV without
scram intervention is low.
Since the scram function and the current margins to trip
avoidance are maintained with revised setpoints, there is no
significant increase in the probability or consequences of an
accident previously evaluated.
D. The ATWS reactor vessel steam dome high pressure
recirculation pump trip (RPT) allowable value is raised 80 psi. The
allowable value appears in Section 3.3.4.2, SR 3.3.4.2.3, in the
Unit 1 and Unit 2 Technical Specifications.
Evaluation
The ATWS-RPT high pressure setpoint initiates a trip of the
recirculation pumps, thereby adding negative reactivity following
events in which a scram does not (but should) occur. Section 5.1.3.2
of NEDC-32405P discusses this function in detail.
The current analytical limit for the ATWS-RPT high pressure trip
is 1150 psig. This value was increased 30 psi in the power uprate
ATWS safety evaluations to account for the 30 psi increase in vessel
operating pressure, SRV setpoints, etc. The current allowable value
in the Technical Specifications is 1095 psig. This allowable value
was not set by the current analytical limit, but by the range of the
installed pressure instruments. As part of the power uprate plant
changes, these pressure instruments will be replaced to accommodate
higher pressure, and the allowable value, in conjunction with the
analytical limit used in the safety analysis, will be increased.
Sections 5.1 and 9.3 of NEDC-32405P show the system can
adequately perform its ATWS function with the new setpoint.
Therefore, the proposed change does not cause a significant increase
in the probability or consequences of an accident previously
evaluated.
E. The low-low set (LLS) SRV arming pressure allowable value is
increased 31 psi, consistent with the increase in operating pressure
and high pressure scram allowable value. The LLS arming pressure
allowable value appears in Section 3.3.6.3, Table 3.3.6.3-1,
Function 1, in the Unit 1 and Unit 2 Technical Specifications.
Evaluation
The allowable value for the LLS SRV high pressure arming
setpoint is increased because the high pressure scram setpoint is
increased. No changes to the LLS arming logic associated with the
SRV tailpipe pressure switches and the LLS opening and closing
pressure setpoints are proposed.
The LLS relief logic mitigates the postulated containment loads
of subsequent SRV actuations during small or intermediate loss of
coolant accidents (LOCAs) by extending the time between actuations.
The LLS logic requires two separate signals to arm itself for
operation. Specifically, the LLS logic arms when an SRV opens (i.e.,
tailpipe pressure switch) and reactor pressure concurrently exceeds
the scram setpoint. To preserve the hierarchy of pressure setpoints,
the high pressure input to the LLS SRV arming logic has the same
setpoint as the high pressure scram, thus minimizing the potential
for a spurious SRV opening through the LLS logic without occurrence
of a reactor scram.
Increasing the arming setpoint is consistent with increasing the
high pressure scram setpoint and will not increase the probability
or consequences of an accident previously evaluated.
F. Lower the permissible rod line for single-loop operation
(SLO) below 45 percent core flow from the 80 percent rod line to the
76 percent rod line. This Technical Specifications limit appears in
Section 3.4.1 (Figure 3.4.1-1) and the corresponding Bases Section B
3.4.1 of the Unit 1 and Unit 2 Technical Specifications.
Evaluation
During development of the generic power uprate program, GE and
the NRC agreed to maintain the current exclusion region in the
power-to-flow map related to thermal-hydraulic stability. The
current limit for SLO is the 80 percent rod line. Power uprate will
redefine 100 percent rated power and, therefore, rated rod or flow
control lines. The 76 percent rod line at uprated conditions closely
corresponds on an absolute, rather than percentage basis, to the
existing 80 percent rod line.
Therefore, this proposed Technical Specifications change ensures
that power uprate operation will not cause a significant increase in
the probability or consequences of accident previously evaluated.
G. The SRV lift setpoints in the Units 1 and 2 Technical
Specifications SR 3.4.3.1 will be increased 30 psi.
Evaluation
The SRVs are designed to prevent overpressurization of the
reactor pressure vessel during abnormal operational transients. The
SRV lift setpoints are increased to accommodate the increase in
operating pressure that accompanies power uprate. The increase in
SRV setpoints ensures that adequate margins are maintained so that
the increase in dome pressure during normal operation does not
result in an increase in the number of unnecessary SRV actuations.
The setpoint increase also maintains the hierarchy of pressure
setpoints described in these proposed changes. Transient evaluations
include a +3 percent tolerance to the nominal setpoints. As
described in Section 3.2 of NEDC-32405P, peak vessel pressure
increases by 3 percent, but remains well below the 1375 psig ASME
Code limit.
Although not credited in the transient analysis, GPC installed a
pressure transmitter system which can electronically actuate the
SRVs on high vessel pressure. The nominal trip setpoints for its
actuation correspond with the nominal mechanical lift setpoints in
the Technical Specifications. The SRV pressure transmitter system
nominal setpoints will also be increased 30 psi.
General Electric generically evaluated the adequacy of BWR SRVs
to operate at uprated temperatures and pressures. The reactor
operating pressure and temperature increases of less than 40 psi and
5 deg.F, respectively, used in that evaluation bound the uprated
Hatch operating conditions.
The impact of power uprate on the Hatch containment dynamic
loads due to SRV discharge has also been evaluated. As discussed in
Section 4.1.2 of NEDC-32405P, the vent thrust loads with power
uprate were calculated to be less than the loads used in the
containment analysis. The effects of power uprate on SRV air-
clearing, the
[[Page 35074]]
discharge line, the pool pressure boundary, and submerged structure
drag loads are discussed in Section 4.1.2 of NEDC-32405P which
concludes that the small increase in the setpoint pressure is well
within the margin in the SRV loads defined in the Mark I Containment
Long-Term Program. Therefore, power uprate does not impact the Hatch
SRV load definitions used in the containment analysis, and no
significant increase in the probability or consequences of an
accident previously evaluated is caused by this proposed change.
H. The Limiting Condition for Operation (LCO) and SRs for the
maximum reactor steam dome pressure will be increased from 1020 psig
to 1058 psig. This requirement appears in LCO 3.4.10, SR 3.4.10.1,
and the corresponding Bases in the Unit 1 and Unit 2 Technical
Specifications.
Evaluation
As discussed in the Technical Specifications Bases and NEDC-
32405P, the maximum reactor dome pressure is an initial condition of
the vessel overpressure protection analysis, which assumes a fast
isolation of all four main steam lines by the main steam isolation
valves (MSIVs). The reactor scram signal generated directly by the
valve closure is assumed defeated for this analysis. Instead, the
scram signal is generated by high neutron flux. The overpressure
analysis for power uprate assumed an initial dome pressure of 1058
psig, which represents an increase of 38 psig. This initial pressure
was chosen approximately 2 percent above the 1035 psig steam dome
operating pressure expected for power uprate operation. The analysis
also included the other changes (including SRV setpoints) discussed
in these proposed changes. Therefore, there is no significant
increase in the probability or consequences of an accident
previously evaluated.
I. The HPCI and RCIC surveillance test pressures in Units 1 and
2 Technical Specifications SRs 3.5.1.8 and 3.5.3.3, respectively,
are increased 38 psi.
Evaluation
The allowable HPCI and RCIC surveillance test pressure is
increased to correspond with the increase in normal reactor
operating pressure and LCO/SR on maximum reactor pressure that
accompanies power uprate. (As discussed in Item H above, the LCO on
reactor steam dome pressure is increased 38 psi.) The change is
needed to ensure that pressure and power reductions are not required
to perform surveillance testing. The requested changes will allow
the quarterly demonstration of the HPCI and RCIC systems' capability
to perform at normal reactor operating pressures, which meets the
original intent of the Technical Specifications.
The HPCI and RCIC systems have been evaluated and demonstrated
to be capable of injecting design flow rate at the higher reactor
pressure as discussed in Sections 4.2 and 3.8 of NEDC-32405P and in
Reference 2.
Therefore, these changes will ensure that power uprate operation
will not cause a significant increase in the probability or
consequences of an accident previously evaluated.
J. Bases Changes
Several changes to the Hatch Units 1 and 2 Technical
Specifications Bases are proposed for consistency with the power
uprate safety analyses. These proposed changes are in addition to
the Bases changes corresponding to proposed changes A through I.
i. The main steam line flow differential pressure setpoints
(Bases Section B 3.3.6.1.c) and the HPCI/RCIC high flow differential
pressure setpoints (Bases Section B 3.3.6.3.a and B 3.3.6.4.a) are
changed for both units.
The allowable values (in percent of rated) will not change for
power uprate operation. However, the actual differential pressure
will change due to the increase in steam flow and pressure.
ii. The HPCI and RCIC upper design pressure in Bases Sections B
3.5.1 and B 3.5.3, respectively, is increased 34 psi for both units
The Bases changes support the design of these high pressure
systems to pump rated flow from approximately 150 psig up to a
pressure associated with the first group of SRV setpoints. This
proposed design pressure conservatively considers the 30 psi higher
nominal setpoints and 3 percent setpoint drift. The capability of
the HPCI and RCIC systems to deliver design flows at these pressures
is discussed in Reference 2, and was reviewed by GE for the Unit 1
and Unit 2 systems.
Note that the upper design pressure for HPCI and RCIC is
different from the surveillance test pressure for HPCI and RCIC
discussed previously in item I. The maximum surveillance test
pressure corresponds to reactor operating pressure, since the
surveillance test is performed when the unit is operating. The HPCI
and RCIC upper design pressure reflects the capability to inject
water to the vessel following a reactor scram and isolation.
iii. The peak post accident containment pressure (Pa) is
changed to 49.6 psig (Unit 1) and 45.5 psig (Unit 2). These values
appear in Bases Sections B 3.6.1.1, B 3.6.1.2, and B 3.6.1.4 in each
unit's Technical Specifications.
Section 4.1.1.3 of NEDC-32405P discusses the peak short-term
containment pressure response which was recalculated for power
uprate conditions. Containment pressure and temperatures remain
below design limits and are essentially unchanged.
iv. The main condenser offgas gross gamma activity rate limit of
240 mci/second will not be changed for power uprate. A statement
that the current limit is conservative for power uprate conditions
was added to Bases Section 3.7.6 for both units.
The Bases derive the current 240 mci/second limit using a rated
core thermal power limit of 2436 MWt. A slightly higher limit could
be justified using the uprated power level. However, adequate margin
exists with the current limit.
v. The inservice hydrostatic and leak testing pressures shown in
Bases Section 3.10.1 are increased 33 psi and 30 psi, respectively.
This change affects each unit's Bases.
This change is a direct result of the 30 psi increase in normal
operating pressure proposed for power uprate. The leakage test is
normally performed at operating pressure and the hydrostatic test at
approximately 110 percent of operating pressure.
The above Bases changes Items i-v have been evaluated and will
not increase the probability or consequences of an accident
previously evaluated.
2. Will the changes create the possibility of a new or different
kind of accident from any accident previously evaluated?
Evaluation
The Operating License changes in power level and the associated
Technical Specifications changes discussed previously will not
create the possibility of a new or different kind of accident from
any accident previously evaluated, as summarized below.
Equipment that could be affected by power uprate was evaluated.
No new operating mode, safety-related equipment lineup, accident
scenario, or equipment failure mode were identified. The full
spectrum of accident considerations defined in RG 1.70 was
evaluated, and no new or different kind of accident was identified.
Uprate uses already-developed technology and applies it within the
capabilities of existing plant equipment in accordance with
presently existing regulatory criteria to include NRC-approved
codes, standards, and methods. GE has designed BWRs of higher power
levels than the uprated power of any of the currently operating BWR
fleet, and no new power dependent accidents have been identified.
The Technical Specifications changes required to implement power
uprate require only minor modifications to the plant's
configuration. All changes were evaluated and found to be
acceptable.
3. Will the changes involve a significant reduction in the
margin of safety?
A. Rated Thermal Power is increased to 2558 MWt on page 3 of the
Unit 1 Operating License, page 4 of the Unit 2 Operating License,
and in Section 1.1 (Definitions) of the Unit 1 and Unit 2 Technical
Specifications.
Evaluation
The events analyzed in the FSAR were re-evaluated to demonstrate
that power uprate can be implemented without exceeding any
regulatory limit. Because the applicable safety analysis criteria
and limits are satisfied for power uprate, the margin of safety
associated with the safety limits and other limits identified in the
Technical Specifications will be maintained.
As discussed in NEDC-32405P, the safety margins prescribed by
the Code of Federal Regulations are maintained by meeting the
appropriate regulatory criteria. Similarly, the margins provided by
the application of the ASME design criteria are maintained. Section
11.4.2 of NEDC-32405P discusses the effects of power uprate on
safety margins for the following:
Fuel thermal limits Design basis accidents and the challenges to
fuel, containment, and radiological releases. Transient analyses.
Non-LOCA radiological releases. Environmental consequences.
These evaluations conclude that applicable safety analysis
criteria and limits are
[[Page 35075]]
satisfied, and thus, the margin of safety will not be significantly
reduced.
B. The surveillance test discharge pressure for the SLC pump at
41.2 gpm is increased from 1190 psig to 1201 psig. This value
appears in SR 3.1.7.7 and corresponding Bases Section B 3.1.7 in the
Unit 1 and Unit 2 Technical Specifications.
Evaluation
Power uprate operation will result in a 30 psi increase in
reactor operating pressure. Several pressure-dependent setpoints
(including SRV setpoints) will be increased to preserve current
margins. Increasing the pressure 11 psi, at which a 41.2 gpm flow
rate is developed, assures continued conformance to ATWS criteria at
uprated conditions. The surveillance test pressure is based on the
maximum pressure for an ATWS event during the time period when the
SLC pump is in operation. Section 6.5 of NEDC-32405P discusses the
capability of these positive displacement pumps. A small increase in
the SRV setpoints will have no effect on the rated injection flow to
the reactor.
For power uprate, the capability of the SLCS to respond with
adequate margin to an ATWS event was confirmed. The results are
reported in Section 9.3.1 of NEDC-32405P. The limiting ATWS event
was an inadvertent MSIV closure. The event was reanalyzed at uprate
conditions with the higher SRV setpoints and ATWS-RPT setpoints.
Peak vessel pressure was well below the ASME emergency limit of 1500
psig. The effect of power uprate on peak clad temperature and
maximum suppression pool temperature was judged to be negligible,
because the calculations showed no increase in fuel surface heat
flux or integrated SRV flow.
In summary, all ATWS criteria are satisfied and the SLC pumps
are capable of injecting the required amounts of sodium pentaborate
at uprated conditions. Therefore, there is no significant decrease
in the margin of safety.
C. The reactor vessel steam dome high pressure allowable value
for RPS instrumentation is increased 31 psi, consistent with the
nominal pressure increase for power uprate. The allowable value
appears in Section 3.3.1.1, Table 3.3.1.1-1, Function 3, in the Unit
1 and Unit 2 Technical Specifications.
Evaluation
The reactor vessel steam dome high pressure scram limit is
increased because the steam dome operating pressure is increased.
Operating pressure for uprated power is increased to assure that
satisfactory reactor pressure control is maintained. The operating
pressure was chosen on the basis of steam line pressure drop
characteristics and the steam flow capability of the turbine.
Satisfactory reactor pressure control requires an adequate flow
margin between the uprated operating condition and the steam flow
capability of the turbine control valves at maximum stroke. An
operating dome pressure of 1035 psig, which is 30 psi higher than
the current operating dome pressure, is expected. Therefore, the
high pressure scram is increased approximately the same amount to
preserve existing margins to reactor trips.
The increases in the steam dome high pressure scram instrument
setpoints for uprated power were evaluated by determining whether
the high pressure scram, which is used as a backup to other scram
signals, provides adequate overpressure protection. The evaluation
demonstrates that the backup protection function, with the revised
setpoints, continues to provide adequate overpressure protection at
uprated power conditions by meeting the applicable ASME Code
criteria. Therefore, there is no significant decrease in the margin
of safety.
D. The ATWS reactor vessel steam dome high pressure RPT
allowable value is raised 80 psi. The allowable value appears in
Section 3.3.4.2, SR 3.3.4.2.3, in the Unit 1 and Unit 2 Technical
Specifications.
Evaluation
The ATWS-RPT high pressure setpoint initiates a trip of the
recirculation pumps, thereby adding negative reactivity following
events in which a scram does not (but should) occur. Section 5.1.3.2
of NEDC-32405P discusses this function in detail.
For power uprate, the capability of the SLCS to respond to a
postulated ATWS event with adequate margin was confirmed (Section
9.3.1 of NEDC-32405P). By reducing reactor power until the SLCS can
inject the required amounts of sodium pentoborate to achieve full
shutdown, the RPT also reduces suppression pool temperature for
isolation cases (also shown to be acceptable for power uprate
conditions in Section 9.3.1 of NEDC-32405P). Therefore, there is no
significant decrease in a margin of safety.
E. The LLS SRV arming pressure allowable value is increased 31
psi, consistent with the increase in operating pressure and high
pressure scram allowable value. The LLS arming pressure allowable
value appears in Section 3.3.6.3, Table 3.3.6.3-1, Function 1, in
the Unit 1 and Unit 2 Technical Specifications.
Evaluation
The allowable value for the LLS SRV high pressure arming
setpoint is increased, because the high pressure scram setpoint is
increased. No changes to the LLS arming logic associated with the
SRV tailpipe pressure switches, and the LLS opening and closing
pressure setpoints are proposed.
Since this proposed change only affects one of two arming
signals for LLS, the safety analyses are not affected; therefore,
there is not a significant change in the margin of safety.
F. Lower the permissible rod line for SLO below 45 percent core
flow from the 80 percent rod line to the 76 percent rod line. This
Technical Specifications limit appears in Section 3.4.1 (Figure
3.4.1-1) and corresponding Bases Section B 3.4.1 of the Unit 1 and
Unit 2 Technical Specifications.
Evaluation
This change to the power versus flow map restricted zone is made
to maintain the same operating constraints and stability margin that
were established for the current power level. This change avoids any
increase in the possibility of occurrence or any increase in the
potential effects of power oscillations. Therefore, there is no
significant decrease in a margin of safety.
G. The SRV lift setpoints in Surveillance Requirement 3.4.3.1
(both units) will be increased 30 psi.
Evaluation
The SRVs are designed to prevent overpressurization of the
reactor pressure vessel during abnormal operational transients. The
SRV lift setpoints are increased to accommodate the increase in
operating pressure that accompanies power uprate. The increase in
SRV setpoints ensures that adequate margins are maintained so that
the increase in dome pressure during normal operation does not
result in an increase in the number of unnecessary SRV actuations.
The setpoint increase also maintains the hierarchy of pressure
setpoints described in these proposed changes. Transient evaluations
include a + 3 percent tolerance to the nominal setpoints. As
described in Section 3.2 of NEDC-32405P, peak vessel pressure
increases by 3 percent but remains well below the 1375 psig ASME
Code limit. Therefore, there is no significant decrease in the
margin of safety.
H. The Limiting Condition for Operation (LCO) and Surveillance
Requirements for the maximum reactor steam dome pressure will be
increased from 1020 psig to 1058 psig. This requirement appears in
LCO 3.4.10, SR 3.4.10.1, and the corresponding Bases in the Unit 1
and Unit 2 Technical Specifications.
Evaluation
As discussed in the Technical Specifications Bases and in
Section 3.2 of NEDC-32405P, the maximum reactor dome pressure is an
initial condition of the vessel overpressure protection analysis,
which assumes a fast isolation of all four main steam lines by the
main steam isolation valves. It is also used as a sensitivity study
parameter for certain transient and LOCA events.
With this revised limit, peak vessel pressure remains below ASME
Code criteria, transient limits are maintained, and LOCA fuel
performance satisfies the requirements of 10 CFR 50.46 and 10 CFR
50, Appendix K. Therefore, there is no significant decrease in a
margin of safety.
I. The HPCI and RCIC surveillance test pressures in SRs 3.5.1.8
and 3.5.3.3, respectively, (both units) are increased 38 psi.
Evaluation
The allowable HPCI and RCIC surveillance test pressure is
increased to correspond with the increase in normal reactor
operating pressure and LCO/SR on maximum reactor pressure that
accompanies power uprate. (As discussed previously, the LCO on
reactor steam dome pressure is increased 38 psi.)
The purpose of the HPCI and RCIC surveillance test is to provide
periodic demonstration of the systems' ability to perform consistent
with the requirements of the analyses at the higher operating
pressure associated with power uprate conditions. An evaluation of
the HPCI and RCIC systems confirmed their ability to operate at
slightly higher turbine speed and provide design flow
[[Page 35076]]
at power uprate conditions. System performance will be confirmed during
the initial power ascension to uprated conditions (and periodically
thereafter per the Technical Specifications). Therefore, there is no
significant decrease in the margin of safety.
J. Bases Changes
Several changes to the Hatch Units 1 and 2 Technical
Specifications Bases are proposed for consistency with the power
uprate safety analyses. These proposed changes are in addition to
the Bases changes corresponding to proposed changes A through I.
i. The main steam line flow differential pressure setpoints, as
shown in Bases Section B 3.3.6.1.c, and the HPCI/RCIC high flow
differential pressure setpoints (Units 1 and 2 Bases Sections B
3.3.6.3.a and B 3.3.6.4.a) are changed.
The allowable values (in percent of rated) will not change for
power uprate operation. However, the actual differential pressure
will change due to the increase in steam flow and pressure.
ii. The HPCI and RCIC upper design pressure in Units 1 and 2
Bases Sections B 3.5.1 and B 3.5.3, respectively, is increased 34
psi.
The Bases changes support the design of these high pressure
systems to pump rated flow from approximately 150 psig up to a
pressure associated with the first group of SRV setpoints. This
proposed design pressure conservatively considers the 30 psi higher
nominal setpoints and 3 percent setpoint drift. The capability of
the Unit 1 and Unit 2 HPCI and RCIC systems to deliver design flows
at these pressures was reviewed by GE and is discussed in Reference
2.
iii. The peak post accident containment pressure (Pa) is
changed to 49.6 psig (Unit 1) and 45.5 psig (Unit 2). These values
appear in Units 1 and 2 Bases Sections B 3.6.1.1, B 3.6.1.2, and B
3.6.1.4.
Section 4.1.1.3 of NEDC-32405P discusses the peak short-term
containment pressure response which was recalculated for power
uprate conditions. Containment pressure and temperatures remain
below design limits and are essentially unchanged.
iv. The main condenser offgas gross gamma activity rate limit of
240 mci/second will not be changed for power uprate. A statement
that the current limit is conservative for power uprate conditions
was added to Units 1 and 2 Bases Section 3.7.6.
The Bases derive the current 240 mci/second limit using a rated
core thermal power limit of 2436 MWt. A slightly higher limit could
be justified using the uprated power level. However, adequate margin
exists with the current limit.
v. The inservice hydrostatic and leak testing pressures shown in
Units 1 and 2 Bases Section 3.10.1 are increased 33 psi and 30 psi,
respectively.
This change is a direct result of the 30 psi increase in normal
operating pressure proposed for power uprate. The leakage test is
normally performed at operating pressure and the hydrostatic test at
approximately 110 percent of operating pressure.
The above Bases changes i-v were evaluated, and there is no
significant decrease in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC
20037.
NRC Project Director: Herbert N. Berkow.
Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric
Authority of Georgia, City of Dalton, Georgia, Docket No. 50-366, Edwin
I. Hatch Nuclear Plant, Unit 2, Appling County, Georgia
Date of amendment request: April 14, 1995.
Description of amendment request: The licensee proposes to revise
Plant Hatch Unit 2 Technical Specifications (TS) to eliminate selected
response time testing requirements from the TS. Specifically, the
response time testing to be eliminated includes sensors and specified
loop instrumentation for: (1) the Reactor Protection System, (2) the
Isolation System, and (3) the Emergency Core Cooling System (ECCS). The
deletion of instrumentation from the ECCS response time testing
necessitates moving the remaining portion of the test to the ECCS
system TS. In addition, the Note for Surveillance Requirement
3.3.6.1.7, which reads: ``Radiation detectors may be excluded,'' is
being removed since response time testing is not required for any
radiation detector that provides a primary containment isolation signal
as indicated in Table 3.3.6.1-1.
Proposed TS Changes 1, 2, and 3 are supported by an analysis
performed by the BWR Owners' Group (BWROG), with the licensee's
participation. The analysis was submitted to the NRC for approval as
Topical Report NEDO-32291, ``System Analyses for the Elimination of
Selected Response Time Testing Requirements,'' Boiling Water Reactor
Owners' Group, January 1994. The NRC approved the Topical Report by a
Safety Evaluation Report (SER) issued on December 28, 1994,
``Evaluation of Boiling Water Reactor Owners' Group Topical Report
NEDO-32291, System Analyses for the Elimination of Selected Response
Time Testing Requirements.'' The BWROG analysis demonstrates that other
periodic tests required by TS, such as channel calibrations, channel
checks, channel functional tests, and logic system functional tests,
ensure that instrument response times are within acceptable limits. The
applicability of the referenced analysis to Plant Hatch has been
verified. Proposed Change 4 removes an unnecessary note, since no
functions subject to this surveillance include radiation monitors.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
Basis for Proposed Changes 1, 2, and 3
1. The changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated. The
purpose of the proposed changes is to eliminate response time
testing requirements for selected instrumentation in the RPS
[Reactor Protection System], Isolation System], and ECCS. However,
because of the continued application of other existing Technical
Specifications requirements, such as channel calibrations, channel
checks, channel functional tests, and logic system functional tests,
the response time of these systems will be maintained within the
acceptance limits assumed in plant safety analyses. This will assure
successful mitigation of an initiating event. The proposed Technical
Specifications changes do not affect the capability of the
associated systems to perform their intended function within their
required response time.
The BWR Owners' Group (BWROG) has documented an evaluation in
NEDO-32291, ``System Analyses for Elimination of Selected Response
Time Testing Requirements,'' which was submitted to the NRC for
review and approval as a Topical Report in January 1994 and
subsequently approved by an NRC SER in December 1994. This
evaluation demonstrates that response time testing is redundant to
the other Technical Specifications requirements listed in the
preceding paragraph. These other tests are sufficient to identify
failure modes or degradation in instrument response time and ensure
operation of the associated systems within acceptance limits. There
are no known failure modes that can be detected by response time
testing that cannot also be detected by the other Technical
Specifications tests.
2. The proposed changes will not create the possibility of a new
or different kind of accident from any accident previously analyzed.
As discussed above, the proposed Technical Specifications changes do
not affect the capability of the associated systems to perform their
intended function within the acceptance limits assumed in plant
safety analyses.
3. The proposed changes do not involve a significant reduction
in the margin of safety. The current Technical Specifications
response times are based on the maximum allowable values assumed in
the plant safety
[[Page 35077]]
analyses, which conservatively establish the margin of safety. As
described above, the proposed Technical Specifications changes do
not affect the capability of the associated systems to perform their
intended function within the allowed response time used as the basis
for the plant safety analyses. Plant and system responses to an
initiating event will remain in compliance with the assumptions of
the safety analyses; therefore, the margin of safety is not
affected.
Although not explicitly evaluated, the proposed Technical
Specifications changes enhance plant safety and operation by:
a. Reducing the time safety systems are unavailable,
b. Reducing safety system actuations,
c. Reducing shutdown risk,
d. Limiting radiation exposure to plant personnel, and
e. Eliminating the diversion of key personnel to conduct
unnecessary testing.
Basis for Proposed Change 4
1. The change does not involve a significant increase in the
probability or consequences of an accident previously evaluated. The
Note for SR 3.3.6.1.7 indicates that response time testing for
radiation detectors that provide primary containment isolation
signals as indicated in Table 3.3.6.1-1 is not required. However,
Table 3.3.6.1-1 does not reference SR 3.3.5.1.7 for any
radiation detector that provides primary containment isolation
signals. The proposed change eliminates the potential for confusion
during instrumentation surveillance testing. Deletion of the note
will not prevent the radiation detectors from performing their
intended function and will not affect the results of any accident
analysis.
2. The proposed changes will not create the possibility of a new
or different kind of accident from any accident previously analyzed.
As discussed above, the proposed Technical Specifications change
eliminates the potential for confusion during instrumentation
surveillance testing. This change does not modify any plant
equipment or change any plant procedure that provides instructions
for the operation of plant equipment. Therefore, the proposed change
will not create the possibility of a new or different kind of
accident from any previously analyzed.
3. The proposed change does not involve a significant reduction
in the margin of safety. The Note that is being deleted by the
change states that testing is not required for instrument sensors
which is not required by the SR. Therefore, the Note is superfluous
and could cause confusion during instrumentation surveillance
testing. The proposed change eliminates that potential. This change
is conservative, since it deletes a statement that was intended to
reduce the amount of surveillance testing performed on certain
instrumentation. The proposed change does not affect plant
equipment, procedures, or radiation release prevention and
mitigating functions. Therefore, the proposed change does not
involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513.
Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric
Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-424 and
50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,
Georgia
Date of amendment request: March 17, 1995.
Description of amendment request: The amendments would revise
Technical Specification (TS) 3.9.4, Containment Building Penetrations,
to allow the personnel airlock to be open during core alterations or
movement of irradiated fuel within the containment.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change to the Technical Specifications does not
involve a significant increase in the probability or consequences of
an accident previously evaluated. The proposed change to
Specification 3.9.4 would allow the containment personnel airlock
(PAL) to be open during fuel movement and core alterations. The PAL
is currently closed during fuel movement and core alterations to
prevent the escape of radioactive material in the event of a fuel
handling accident. The PAL is not an initiator to any accident.
Whether the PAL doors are opened or closed during fuel movement or
core alterations has no effect on the probability of any accident
previously evaluated.
Allowing the PAL doors to be open during fuel movement and core
alterations does increase the consequences of a fuel handling
accident in the containment from essentially no offsite dose release
to an estimated release of 65.6 rem to the thyroid and 0.28 rem to
the whole body. However, the calculated offsite dose release is
lower than the case analyzed in the FSAR [Final Safety Analysis
Report] for an accident in the Spent Fuel Pool, with no filtration
of the resulting release. In addition, the calculated doses are
larger than the expected doses because the calculation does not
incorporate the closing of the PAL door after the containment is
evacuated. Closing the airlock door within 15 minutes results in a
calculated offsite dose of 8.2 rem to the thyroid and 0.025 rem
whole body. The projected dose to control room operators was
reviewed and the projected dose remained below SRP acceptance limits
as long as control room emergency ventilation was established within
7 minutes. It was assumed the individual assigned to close the
airlock doors remained stationed at the airlock for 15 minutes. A
best estimate dose analysis indicated this individual could be
expected to receive 5.6 rem to the thyroid and 0.15 rem whole body.
The proposed change will significantly reduce the dose to other
workers in the containment in the event of a fuel handling accident
by speeding the containment evacuation process. The proposed change
will also significantly decrease the wear on the PAL doors and,
consequently, increase the availability of the PAL doors in the
event of an accident.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change to the Technical Specifications does not
create the possibility of a new or different kind of accident from
any accident previously evaluated because the proposed change
affects a previously evaluated accident, e.g., a fuel handling
accident. It does not represent a significant change in the
configuration or operation of the plant and, therefore, does not
create the possibility of a new or different type of accident from
any accident previously evaluated.
3. The proposed change to the Technical Specifications does not
involve a significant reduction in a margin of safety. The margin of
safety as defined by 10 CFR Part 100 for a fission product release
is 300 rem thyroid and 25 rem whole body for an individual exposed
at the site boundary for two hours. The analysis shows values that
are well below the acceptance limits. In fact, the margin remains
essentially the same as previously evaluated by the NRC. There is no
increase in calculated offsite dose resulting from a fuel handling
accident. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
Based upon the preceding information, it has been determined
that the proposed Technical Specifications addition does not involve
a significant hazards consideration as defined by 10 CFR 50.92.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Burke County Public Library,
412 Fourth Street, Waynesboro, Georgia 30830.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308.
NRC Project Director: Herbert N. Berkow.
[[Page 35078]]
Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric
Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-424 and
50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,
Georgia
Date of amendment request: May 12, 1995.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) to support a one-time
exemption from the requirement of Section III.D.1(a) of 10 CFR Part 50,
Appendix J, and any other future Appendix J exemptions that may be
approved by the NRC for Vogtle, Unit 1. Specifically, the TS change
would insert the words ``Except as modified by NRC approved
exemptions'' at the beginning of the first sentence of TS Surveillance
Requirement 4.6.1.2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The change does not involve a significant increase in the
probability or consequences of an accident previously evaluated. The
proposed change does not involve a change to structures, systems, or
components which would affect the probability of an accident
previously evaluated in the Vogtle Electric Generating Plant (VEGP)
Final Safety Analysis Report (FSAR). The change only provides a
mechanism for implementing exemptions to 10 CFR 50, Appendix J
containment leak rate testing criteria which have been approved by
the NRC.
2. The proposed change will not create the possibility of a new
or different kind of accident from any accident previously analyzed.
The amendment would not change the design, configuration, or method
of plant operation. It only allows exemption to specific 10 CFR 50,
Appendix J criteria as previously approved by the NRC.
3. Operation of VEGP, Unit 1 in accordance with the proposed
change will not involve a significant reduction in the margin of
safety. The proposed change would not, in itself, change a safety
limit, an LCO, or a surveillance requirement on equipment required
for plant operation. Before the change could be used an exemption to
10 CFR 50, Appendix J would have to be evaluated and approved by the
NRC. The change only provides a way to implement NRC approved
exemptions without violating the Technical Specifications.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Burke County Public Library,
412 Fourth Street, Waynesboro, Georgia 30830.
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile Island
Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of amendment request: June 1, 1995.
Description of amendment request: The proposed license amendment
would revise the Technical Specifications (T.S.) for Three Mile Island
Nuclear Station, Unit 1 (TMI-1) to delete the remaining portions of the
TMI-1 Radiological Effluent Technical Specifications (RETS) and
relocate them in accordance with the guidance contained in the Generic
Letter 89-01 (GL 89-01) and NUREG-1430. The proposed change would also
modify the Radiation Monitoring Systems surveillance requirements to
specify only those radiation monitors that have Limiting Conditions for
Operation (LCO), and revise some of the calibration frequencies.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence or the consequences of an accident
previously evaluated. The proposed amendment allows relocation of
the remaining RETS to the ODCM [Offsite Dose Calculation Manual]
according to the guidance contained in GL 89-01 and NUREG-1430. This
proposal simplifies the RETS, meets the regulatory requirements for
radioactive effluent controls and radiological environmental
monitoring, and is provided as a line-item improvement of the T.S.
In addition, this proposed amendment specifies surveillance
requirements only for those radiation monitors that have an LCO or
specified operability requirements. The radiation monitors that are
currently included in the T.S. surveillance program but have no
associated LCO or specified operability requirement will be placed
in the PM [preventive maintenance] program.
Finally, the proposed amendment extends the interval between
successive calibration surveillances for those radiation monitors
evaluated herein. This change does not involve any change to the
actual surveillance requirements, nor does it involve any change to
the limits or restrictions on plant operations. The reliability of
systems and components relied upon to prevent or mitigate the
consequences of accidents previously evaluated is not degraded
beyond that obtained from the currently defined quarterly interval.
Assurance of system and equipment availability is maintained.
This change does not involve any change to system or equipment
configuration. Therefore, this change does not significantly
increase the probability of occurrence or the consequences of an
accident previously evaluated.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposal in part relocates procedural details, currently
included in the T.S., on radioactive effluents to the ODCM. Future
changes to these procedural details in the ODCM will be handled
under the administrative controls for changes to the ODCM.
In addition, this proposed amendment specifies surveillance
requirements only for those radiation monitors that have an LCO or
specified operability requirements. The radiation monitors that are
currently included in the T.S. surveillance program but have no
associated LCO or specified operability requirement will be placed
in the PM program.
Finally, the proposed amendment extends the interval between
successive calibration surveillances for those radiation monitors
evaluated herein. This change does not involve any change to the
actual surveillance requirements, nor does it involve any change to
the limits and restrictions on plant operations. This change does
not involve any change to system or equipment configuration.
Therefore, this change is unrelated to the possibility of
creating a new or different kind of accident from any previously
evaluated.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The procedural details being relocated to the ODCM are
consistent with the guidance provided in GL 89-01 and NUREG-1430.
In addition, this proposed amendment specifies surveillance
requirements only for those radiation monitors that have an LCO or
specified operability requirements. The radiation monitors that are
currently included in the T.S. surveillance program but have no
associated LCO or specified operability requirement will be placed
in the PM program.
Finally, the proposed amendment extends the interval between
successive calibration surveillances for those radiation monitors
evaluated herein. This change does not involve any change to the
actual surveillance requirements, nor does it involve any change to
the limits and restrictions on plant operations. The reliability of
the radiation monitors is not significantly degraded beyond that
obtained from the currently defined surveillance interval. Assurance
of system availability is maintained.
Therefore, it is concluded that operation of the facility in
accordance with the proposed amendment does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this
[[Page 35079]]
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Law/Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Phillip F. McKee.
Gulf States Utilities Company, Cajun Electric Power Cooperative, and
Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit
1, West Feliciana Parish, Louisiana
Date of amendment request: May 30, 1995.
Description of amendment request: The proposed amendment would
revise the technical specifications (TS) to increase the surveillance
test period for the containment integrated leak rate test (ILRT) from
40 plus or minus 10 months to every 10 years based on past performance.
The change would also require testing on a more frequent basis if any
test failures were to occur and to return to the 10 year period with
subsequent performance improvements.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) The proposed change does not involve a change to the plant
design or operation. As a result, the proposed change does not
affect any of the parameters or conditions that contribute to
initiation of any accidents previously evaluated. Thus, the proposed
change cannot increase the probability of any accident previously
evaluated. The proposed change potentially affects the leak tight
integrity of the containment structure designed to mitigate the
consequences of a loss of coolant accident (LOCA). The function of
the containment is to maintain functional integrity during and
following the peak transient pressures and temperatures which result
from any loss of coolant accident (LOCA). The containment is
designed to limit fission product leakage following the design basis
LOCA and analyses demonstrate that these offsite doses are less than
those allowed under 10CFR100 design limits of 15 psig and 185
deg.F. Because the proposed change does not alter the plant design,
only the frequency of measuring containment leakage, the proposed
change does not directly result in an increase in containment
leakage. However, decreasing the test frequency can increase the
probability that a large increase in containment leakage could go
undetected for an extended period of time. These leakage paths
include potential cracks in the containment structure and various
penetrations through the containment structure. Based upon the
results of the structural integrity test conducted as part of the
preoperational or preservice test program and the periodic
containment and drywell structural integrity surveillance tests,
additional cracking of the containment is not expected during the
remaining life to the plant. Ventilation and piping penetrations are
designed with two isolation valves in series with one valve in the
drywell and another either outside primary containment or in the
wetwell. High energy lines that extend into the wetwell, such as the
Main Steam and Feedwater lines, are encapsulated by guard pipes to
direct energy to the drywell in case of a piping rupture.
Electrical penetrations are sealed with a high strength/density
material that will prevent leakage as well as provide radiation
shielding. The TS ILRT acceptance criterion of 0.75 La [maximum
allowable leakage rate at the calculated maximum accident pressure,
Pa] provides margin for degradation. Containment performance
data to date suggests that containment degradation, even during a
ten (10) year interval between tests, will not exceed this margin.
Based on the above, EOI [Entergy Operations, Inc.] has concluded
that the proposed change will not result in a significant increase
in the probability or consequences of any accident previously
evaluated.
(2) The proposed change does not involve a change to the plant
design or operation. As a result, the proposed change does not
affect any of the parameters or conditions that could contribute to
initiation of any accidents. This change involves the reduction in
the Integrated Leak Rate Test frequency. The method of performing
the test is not changed. No new accident modes are created by
extending the testing intervals. No safety-related equipment or
safety functions are altered as a result of this change. Extending
the test frequency has no influence on, nor does it contribute to,
the possibility of a new or different kind of accident or
malfunction from those previously analyzed. Based upon the above,
EOI has concluded that the proposed change will not create the
possibility or a new or different kind of accident previously
evaluated.
(3) The proposed change only affects the frequency of measuring
containment leakage and does not change the leakage rate limit.
However, the proposed change can increase the probability that a
large increase in containment leakage could go undetected for an
extended period of time. Operational experience has shown that the
leak tightness of the containment has been maintained significantly
below the allowable leakage limit. In fact, an analysis was
conducted to determine the potential risk to the public from the
proposed change. Based on this analysis, under several different
accident scenarios, the risk of radioactivity release from
containment was found to be negligible.
The margin of safety that has the potential of being impacted by
the proposed change involves the offsite dose consequences of
postulated accidents which are directly related to containment
leakage rate. The containment isolation system is designed to limit
leakage to La which is defined by the RBS Technical
Specifications to be 0.26 percent by weight of the containment air
per 24 hours at 7.6 psig (Pa). The limitation on containment
leakage rate is designed to ensure that total leakage volume will
not exceed the value assumed in the accident analyses at the peak
accident pressure (Pa) or 7.6 psig.
To provide additional conservatism, the measured overall
integrated leakage rate is further limited to less than or equal to
0.75 La during performance of the periodic Integrated Leak Rate
Test and to less than or equal to 0.60 La (total combined
leakage) for Type B and C leak rate tests. This is done to account
for the possible degradation of the containment leakage barriers
between tests. These acceptance criteria ensure that an acceptable
margin of safety is being maintained and will not be altered by the
proposed change. The preservation of this margin will continue to
provide for potential degradation of the leakage barriers between
tests. RBS [River Bend Station] presently has on docket with the
staff a submittal (reference RBG-41133, Rev. 1 to LAR 93-14 dated
January 18, 1995) that allows the acceptance criteria, between
required leakage rate tests, to be less than or equal to 1.0 La
since at less than or equal to 1.0 La, the offsite does
consequences are bounded by the assumptions of safety analysis.
No change in the method of testing is being proposed. The Type A
test will continue to be done at full pressure (Pa) or greater.
Primary containment penetrations which require Type B or C leak
tests will be performed in the same manner as before. Other programs
are in place to ensure that proper maintenance and repairs are
performed during the service life of the primary containment and
systems and components penetrating the primary containment.
No change in the RBS allowable leakage rate is being proposed.
These conservative leakage rates ensure that the containment leakage
remains low. As a result, EOI has concluded that the proposed change
will not result in a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, Louisiana 70803.
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
[[Page 35080]]
1400 L Street, N.W., Washington, DC 20005.
NRC Project Director: William D. Beckner.
Gulf States Utilities Company, Cajun Electric Power Cooperative, and
Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit
1, West Feliciana Parish, Louisiana
Date of amendment request: May 30, 1995.
Description of amendment request: The proposed amendment would
revise the technical specifications (TS) to increase the time period
for drywell leakage tests from eighteen months to five years based on
performance. The new surveillance requirements would also reduce the
time period if any failures occur and limit subsequent periods until
drywell leakage test performance again improves.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) The proposed change does not involve a change to the plant
design or operation. As a result, the proposed change does not
affect any of the parameters or conditions that contribute to
initiation of any accidents previously evaluated. Thus, the proposed
change cannot increase the probability of any accident previously
evaluated.
The proposed change potentially affects the leak tight integrity
of the drywell, a structure used to mitigate the consequences of a
loss of coolant accident (LOCA). The function of the drywell is to
channel the steam released from a LOCA through the suppression pool,
limiting the amount of steam released to the primary containment
atmosphere. This limits the containment pressurizations due to the
LOCA. The leakage of the drywell is limited to ensure that the
primary containment does not exceed its design limits of 185 deg.F
and 15 psig. Because the proposed change does not alter the plant
design, only the frequency of measuring the drywell leakage, the
proposed change does not directly result in an increase in drywell
leakage. However, decreasing the test frequency can increase the
probability that a large increase in drywell bypass leakage could go
undetected for an extended period of time. There are several
potential sources of steam bypass leakage paths. These include
potential cracks in the drywell concrete structure and various
penetrations through the drywell structure. Based upon the results
of the structural integrity test conducted as part of the
preoperational or preservice test program, additional cracking of
the drywell is not expected during the remaining life of the plant.
Ventilation and piping penetrations are designed with two isolation
valves in series with one valve in the drywell and another either
outside primary containment or in the wetwell. High energy lines
that extend into the wetwell, such as the Main Steam line and
Feedwater lines, are encapsulated by guard pipe to direct energy to
the drywell in case of a piping rupture. Electrical penetrations are
sealed with a high strength/density material that will prevent
leakage as well a provide radiation shielding. The TS DBLRT [Drywell
Bypass Leakage Rate Tests] acceptance criterion of 10% of the design
bypass leakage area parameter provides margin for degradation.
Drywell performance data to date suggests that drywell degradation,
even during a five year interval between tests, will not exceed this
margin. RBS presently has on docket with the staff a submittal
(reference EOI letter RBG-41133, Rev. 1 to LAR 93-14 dated January
18, 1995) that allows the acceptance criteria, between required
leakage rate tests, to be (bypass leakage area parameter) since at
(bypass leakage area parameter) the containment temperature and
pressurization response are bounded by the assumptions of the safety
analysis.
Based on the above, EOI has concluded that the proposed change
will not result in a significant increase in the consequences of any
accident previously evaluated.
(2) The proposed change does not involve a change to the plant
design or operation. As a result, the proposed change does not
affect any of the parameters or conditions that could contribute to
initiation of any accidents. Thus, the proposed change cannot create
the possibility of an accident not previously evaluated.
(3) The proposed change only affects the frequency of measuring
the drywell bypass leakage rate and does not change the bypass
leakage limit for the drywell. However, the proposed change can
increase the probability that a large increase in drywell bypass
leakage could go undetected for an extended period of time.
Operational experience has shown that the leak tightness of the
drywell has been maintained significantly below the allowable
leakage limits. In fact, an analysis was conducted to determine the
potential risk to the public from the proposed change. Based on this
analysis, under several different accident scenarios, the risk of
radioactivity release from containment was found to be negligible.
As a result, EOI has concluded that the proposed change will not
result in a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, Louisiana 70803.
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005.
NRC Project Director: William D. Beckner.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, Michigan
Date of amendment requests: May 25, 1995 (AEP:NRC:107IT).
Description of amendment requests: The proposed amendments would
implement a cycle- and burnup-dependent peaking factor penalty to the
allowable power level. The Technical Specifications would be changed to
refer to the Core Operating Limits Report for this burnup-dependent
penalty.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Per 10 CFR 50.92, a proposed amendment will not involve a
significant hazards consideration if the proposed amendment does
not:
(1) involve a significant increase in the probability or
consequences of an accident previously evaluated,
(2) create the possibility of a new or different kind of
accident from any accident previously evaluated, or
(3) involve a significant reduction in a margin of safety.
Criterion 1
The proposed changes will not involve a significant increase in
the probability of an accident previously evaluated because the
changes will not result in a change to any of the process variables
that might initiate an accident. There are no physical changes to
the plant associated with this T/S change. The consequences of an
accident previously evaluated will not be increased because the
changes increase the penalty applied to FQ when it is measured
to be increasing. FQ and allowable power level (APL) T/S
surveillance requirements are not being changed. Furthermore,
allowing a cycle and burnup dependent FQ penalty to be located
in the COLR was accepted by the NRC in a [November 26, 1993] safety
evaluation on WCAP-10216-P, Rev. 1 [``Relaxation of Constant Axial
Offset Control- FQ Surveillance Technical Specification''].
Criterion 2
The proposed changes will not create the possibility of a new or
different kind of accident from any accident previously evaluated
because the changes will involve no physical changes to the plant
nor any changes in plant operations. Furthermore, the FQ and
APL T/S surveillance requirements are not being changed, and the
change to the FQ penalty is conservative.
Criterion 3
The proposed amendment[s] will not involve a significant
reduction in a margin of safety. When the increased FQ penalty
is applied, it reduces the allowable power level, thus increasing
the margin of safety.
[[Page 35081]]
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Project Director: Cynthia A. Carpenter, Acting.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, Michigan
Date of amendment requests: May 25, 1995 (AEP:NRC:1124B).
Description of amendment requests: The proposed amendments would
modify the Technical Specifications (TS) to allow fuel reconstitution.
The proposed change is a TS line item improvement per NRC Generic
Letter 90-02, supplement 1, ``Alternative Requirements for Fuel
Assemblies in the Design Features Section of Technical
Specifications.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Per 10 CFR 50.92, a proposed change does not involve significant
hazards consideration if the change does not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated,
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated, or
3. Involve a significant reduction in a margin of safety.
Criterion 1
The proposed changes only modify the T/Ss such that
reconstitution is recognized as acceptable under very limited
circumstances. Reconstitution is limited to substitution of
zirconium alloy or stainless steel filler rods, and must be in
accordance with approved applications of fuel rod configurations.
Although these changes permit reconstitution to occur without the
need for a specific T/S change, an approved methodology is required
prior to its application. Since the changes will allow substitution
of filler rods for leaking or potentially leaking rods, the changes
may actually reduce the radiological consequences of an accident. It
is noted that the specific changes requested in this letter have
previously been found acceptable by the NRC in GL 90-02 supplement
1. For these reasons, we conclude that the changes will not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
Criterion 2
The proposed changes will not create the possibility of a new or
different kind of accident from any accident previously evaluated
because they will only affect the assembly configuration and can
only be implemented in accordance with an NRC-approved methodology.
The other aspects of plant design, operation limitations, and
responses to events will remain unchanged. It is noted that the
changes have previously been determined acceptable by the NRC in GL
90-02 supplement 1.
Criterion 3
The proposed amendment will not involve a significant reduction
in a margin of safety because the changes can only be implemented in
accordance with an NRC-approved methodology. It is noted that the
changes have previously been determined acceptable by the NRC in GL
90-02 supplement 1.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Project Director: Cynthia A. Carpenter, Acting.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, Michigan
Date of amendment requests: May 25, 1995 (AEP:NRC:1200B).
Description of amendment requests: The proposed amendments would
modify the Technical Specifications to change the surveillance
frequency of the manual actuation function for main steam line
isolation. This change is consistent with the testing requirements for
associated valves as specified in the American Society of Mechanical
Engineers (ASME) Code Section XI inservice testing program at Cook.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Per 10 CFR 50.92, a proposed change does not involve significant
hazards consideration if the change does not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated,
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated, or
3. Involve a significant reduction in a margin of safety.
Criterion 1
This change will reduce the frequency of the surveillance
testing on the MSIV [main steamline isolation valve] manual
actuation circuitry from monthly to quarterly. Because of the risks
involved in testing the dump valves, the reduction in test frequency
may reduce the probability of an accidental unit trip and valve seat
failure due to repeated cycling. Our review of the surveillance test
history has shown that the system is highly reliable, and gives us
confidence that the change in test frequency will not endanger
public health and safety. Furthermore, the change to a quarterly
surveillance interval is consistent with the testing performed for
the dump valves per ASME Section XI. For these reasons, it is our
belief that the proposed changes do not involve a significant
increase in the probability or consequences of a previously
evaluated accident.
Criterion 2
The changes will not introduce any new modes of plant operation,
nor will any physical changes to the plant be required. Thus, the
changes should not create the possibility of a new or different kind
of accident from any accident previously analyzed or evaluated.
Criterion 3
This change will reduce the frequency of the surveillance
testing on the MSIV manual actuation circuitry from monthly to
quarterly. Our review of the surveillance test history has shown
that the system is highly reliable, and gives us confidence that the
change in test frequency will not endanger public health and safety.
Furthermore, the change to quarterly surveillance is consistent with
the testing performed for the dump valves per ASME Section XI. For
these reasons, it is our belief that the proposed changes do not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Project Director: Cynthia A. Carpenter, Acting.
[[Page 35082]]
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, Michigan
Date of amendment requests: May 26, 1995 (AEP:NRC:1210).
Description of amendment requests: The proposed amendments would
modify the Reactor Trip System Instrumentation and Engineered Safety
Feature Actuation System Instrumentation sections of the Technical
Specifications (TS) to relocate the tables of response time limits to
the Updated Final Safety Analysis Report (UFSAR). These changes are a
line item improvement of the TS in accordance with NRC Generic Letter
93-08, ``Relocation of Technical Specification Tables of Instrument
Response Time Limits.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Per 10 CFR 50.92, a proposed amendment will not involve a
significant hazards consideration if the proposed amendment does
not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated,
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated, or
(3) Involve a significant reduction in a margin of safety.
Criterion 1
The proposed changes will not involve a significant increase in
the probability of an accident previously evaluated because the
changes will not result in a change to any of the process variables
that might initiate an accident. There are no physical changes to
the plant associated with the T/S change. The consequences of an
accident previously evaluated will not be increased because the
changes simply allow relocation of response time limits to the
UFSAR. Time response testing will continue to be required by the T/
Ss. Any changes to the response time values will be made in
accordance with the requirements of 10 CFR 50.59. It is noted that
these T/S changes have previously been determined acceptable by the
NRC in GL 93-08.
Criterion 2
The proposed changes will not create the possibility of a new or
different kind of accident from any accident previously evaluated
because the changes will involve no physical changes to the plant
nor any changes in plant operations. Time response testing will
continue to be required by the T/Ss. Any changes to the time
response values will be made in accordance with the requirements of
10 CFR 50.59. It is noted that these changes have previously been
determined acceptable by the NRC in GL 93-08.
Criterion 3
The proposed amendment will not involve a significant reduction
in a margin of safety because time response testing will continue to
be required by the T/Ss. Any changes to the response time values
will be made in accordance with the requirements of 10 CFR 50.59. It
is noted that these changes have previously been determined
acceptable by the NRC in GL 93-08.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Project Director: Cynthia A. Carpenter, Acting.
North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: May 30, 1995.
Description of amendment request: The proposed amendment would
change the upper limit for the moderator temperature coefficient (MTC)
for certain operating conditions. Specifically, the upper limit
specified in Technical Specification 3.1.1.3 for the MTC would be
changed to +0.5 x 10-4 delta k/k/ deg.F for all rods out at the
beginning of cycle for power levels up to 70% rated thermal power with
a linear ramp to 0 delta k/k/ deg.F at 100% rated thermal power. The
currently specified upper limit for all operating conditions is 0 delta
k/k/ deg.F.
A paragraph would be added to the Basis to Technical Specification
3.1.1.3 providing a commitment to comply with the ATWS Rule and the
basis for the Rule by assuring ATWS core damage frequency will remain
below the Commission established target of 1.0 x 10-5 per reactor
year. The commitment would be implemented by determining a more
restrictive, cycle-specific upper MTC limit and placing it in the Core
Operating Limits Report (COLR).
Additionally, a reference for the analytical method used to
determine the cycle-specific MTC upper limit would be added to TS
6.8.1.6.b.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below.
A. The changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated (10 CFR
50.92(c)(1)). The proposed changes do not affect the manner by which
the facility is operated and do not change any facility design feature
or equipment which influences the initiation of an accident, therefore,
there is no change in the probability of any accident previously
analyzed. Each accident or transient, with the exception of the
Anticipated Transient Without SCRAM (ATWS), has been analyzed for the
proposed changes and has been approved previously by the Commission
with the issuance of Amendment 33 (December 6, 1994) to the Facility
Operating License. The proposed cycle-specific MTC to be included in
the COLR will assure that the consequences of an ATWS will remain
bounded by the analysis previously documented. Therefore, the
consequences of previously evaluated accidents, including ATWS, will
not be significantly increased by the proposed changes.
B. The changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated (10 CFR
50.92(c)(2)) because the changes proposed merely involve changes in the
upper limits of MTC imposed by the Technical Specifications and COLR.
No changes are made to the design or manner of operation of any
structure, system or component and no new failure mechanisms are
introduced.
C. The changes do not involve a significant reduction in a margin
of safety (10 CFR 50.92(c)(3)). The analyses of each accident or
transient previously presented to support the issuance of Amendment 33
were performed using the proposed upper MTC limit, and the results
demonstrated that the acceptance criteria specified for each event are
met. The cycle-specific MTC limit in the COLR will be adjusted to
assure that the acceptance criteria for a postulated ATWS event are met
thereby preserving the margin of safety.
Based on this review, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
[[Page 35083]]
Local Public Document Room location: Exeter Public Library,
Founders Park, Exeter, NH 03833.
Attorney for licensee: Thomas Dignan, Esquire, Ropes & Gray, One
International Place, Boston MA 02110-2624.
NRC Project Director: Phillip F. McKee.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: May 31, 1995.
Description of amendment request: The amendment would provide
additional restrictions on the operation of the component cooling water
(CCW) system heat exchangers to ensure that the CCW system temperature
is maintained within its analyzed design basis.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
In preparation for, and in response to a service water system
operational performance self assessment, the heat loads in the
Component Cooling Water (CCW) system were reevaluated to determine
the peak temperatures on the system and components cooled by the CCW
system. It was determined that if all of the containment coolers
were operating, the return temperature of the CCW system could
exceed the 120 deg.F stated in the Updated Safety Analysis Report
(USAR) as the maximum temperature of the system.
During a Large Break Loss of Coolant Accident (LBLOCA) or a Main
Steam Line Break Inside Containment (MSLB/IC), the containment air
cooling units and containment air cooling and filtering units will
automatically start to remove heat from the containment atmosphere.
The heat sink for the containment air coolers is the CCW system. The
heat removed from the containment atmosphere is transferred to the
Raw Water (RW) system via the component cooling heat exchangers AC-
1A, B, C, and D. The heat is then ultimately rejected to the
Missouri River by the RW system.
Calculations indicate that the CCW return temperature (i.e.,
mixed exit temperature) from the component cooling heat exchangers
could exceed 160 deg.F after a LBLOCA or MSLB/IC with the present TS
minimum requirements for the heat exchangers. Further evaluation
indicated that the CCW system (and components cooled by CCW) could
withstand temperatures above the 120 deg.F temperature stated in the
USAR, but a return temperature above 158 deg.F would require
additional evaluation of thermal-induced stresses on the CCW return
side pipe supports. In order to maintain the peak CCW return
temperature to less than or equal to 158 deg.F, additional
restrictions must be placed on the number of component cooling heat
exchangers required to be operable.
The current minimum requirements for component cooling heat
exchangers are contained in Technical Specification (TS) 2.3,
``Emergency Core Cooling System,'' and require that three of the
four heat exchangers be operable when the plant is in operating
Modes 1 and 2. Analyses show that three in service heat exchangers
will maintain the CCW temperatures in an analyzed range following a
DBA. In order to ensure that three heat exchangers are available, in
conjunction with an assumed single failure, four are required to be
operable. The proposed change would place additional restrictions on
the operation of the CCW heat exchangers by requiring four heat
exchangers to be operable in Modes 1 and 2, and if only three are
operable then provide 14 days to restore the system to four operable
heat exchangers.
The proposed change does not involve a significant increase in
the probability of an accident previously evaluated. The proposed
change does not impact systems, structures, or components that are
initiators of any analyzed accidents.
The proposed change does not involve a significant increase in
the consequences of an accident previously evaluated. The proposed
change ensures that the CCW system and safety-related components
cooled by the CCW will perform their safety functions in response to
previously evaluated accidents. The proposed change was evaluated
utilizing the probabilistic risk analysis model of the FCS
Individual Plant Examination. The IPE concluded that the routine
testing and maintenance activities, for the RW and CCW systems
(e.g., inoperability of components for testing and maintenance) are
not significant contributors to severe accident risk.
Therefore, the proposed change would not increase the
probability or consequences of an accident previously evaluated.
(2) The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change does not create an initiator for a new or
different kind of accident from those previously evaluated. The
proposed change places additional restrictions on the operation of
equipment to ensure that the CCW system and safety-related
components cooled by the CCW will perform their safety functions.
The additional restrictions were evaluated in combination with
existing allowances on RW and CCW pump inoperability, to confirm
that the peak CCW return temperature would be in an analyzed range,
and will not adversely impact the operability of the CCW system or
safety-related components cooled by CCW. These restrictions are
valid up to and including a river temperature of 90 deg.F, which is
the upper bound currently cited in the USAR.
Various single active failures were postulated to determine the
most limiting failure in conjunction with the maximum heat load from
the containment air coolers. It was determined that with the river
temperature less than 70 deg.F, a single failure of a RW valve to
open on a component cooling heat exchanger would not raise the CCW
return temperature to an unanalyzed level, but with the river
temperature greater than or equal to 70 deg.F, the CCW return
temperature could be at an unanalyzed level. Therefore, it is
proposed that when the river temperature is greater than or equal to
70 deg.F four heat exchangers have RW in service (i.e., RW valves
open). Having RW in service eliminates the potential failure of a RW
valve to auto-open as a credible single active failure.
The proposed change ensures that the CCW system and safety-
related components cooled by the CCW will perform their safety
functions. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
(3) The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change provides additional restrictions on the CCW
system and ensures that the CCW system will perform its design
safety function. These additional restrictions ensure that the CCW
system will be capable of removing the maximum heat load from the
containment cooling system following a DBA and thereby ensures that
the containment pressure remains below its limit as assumed in the
USAR. Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102.
Attorney for licensee: James R. Curtiss, Winston & Strawn, 1400 L
Street, N.W., Washington, DC 20005-3502.
NRC Project Director: William H. Bateman.
Pennsylvania Power and Light Company, Docket No. 50-387, Susquehanna
Steam Electric Station, Unit 1, Luzerne County, Pennsylvania
Date of amendment request: May 5, 1995.
Description of amendment request: This amendment would remove from
the Susquehanna Steam Electric Station Unit 2 Technical Specifications,
the listing of three residual heat removal (RHR) system valves in Table
3.6.3-1, ``Primary Containment Isolation Valves'' These valves are no
longer needed to support the steam condensing mode of the RHR system
and are being removed from the plant during the Unit 2 seventh
[[Page 35084]]
refueling and inspection outage in September of this year.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
With the prior deletion of the steam condensing mode of RHR and
the isolation of the high and low pressure interfaces, the three
pressure relief valves that are being removed from the plant have no
active function. Their passive function of maintaining system or
containment integrity will be fulfilled by blind flanges on
equilvent. Also, the RHR and RCIC piping are provided with
overpressure protection from other pressure relief valves.
Therefore, the removal of these pressure relief valves does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The pressure relief valves that are being removed had two
primary functions. First, they provided overpressure protection for
the RHR and RCIC piping during the steam condensing mode of RHR.
Since the steam condensing mode has been deleted from the plant,
these valves no longer have that function. Also, overpressure
protection of the RHR and RCIC piping is provided by other existing
pressure relief valves. Second, these valves maintained system or
containment integrity. When the pressure relief valves are removed
from the plant, they will be replaced with blind flanges or
equivalent that will maintain system or containment integrity.
Therefore, the removal of the three pressure relief valves does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
Since the steam condensing mode of RHR has been eliminated, the
three pressure relief valves have no active function. Their passive
function of maintaining system or containment integrity will be
fulfilled by blind flanges or equivalent. Also, overpressure
protection of RHR and RCIC piping is provided by other existing
pressure relief valves. Therefore, the removal of the three pressure
relief valves does not involve a significant reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701.
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
NRC Project Director: John F. Stolz.
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: May 19, 1995.
Description of amendment request: The proposed Technical
Specifications (TS) change would revise TS Table 3.3.3-3, ``Emergency
Core Cooling System Response Times'' to reflect the value of 60 seconds
for the High Pressure Coolant Injection system response time instead of
30 seconds as currently specified.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed Technical Specifications (TS) change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The proposed TS change will increase the High Pressure Coolant
Injection (HPCI) system response time from 30 seconds to 60 seconds.
The proposed TS change does not involve any physical change in the
plant configuration which may cause an accident, or affect safety-
related equipment performance or cause its failure. There is no
increase in the consequences of an accident, because the HPCI
response time increase does not affect the licensing basis Peak
Cladding Temperature (PCT), which remains below the regulatory limit
of 2200 deg.F.
The Loss of Feedwater Flow (LOFW) event was evaluated for being
potentially affected by the increased HPCI system response time. The
HPCI system is one of the systems which provides reactor vessel
water makeup inventory, and is initiated automatically on a low
reactor water level (Level 2) signal. The LOFW analysis shows that
Level 1 is not reached and that the top of the active fuel will
remain covered throughout the event. Therefore, adequate core
cooling will be maintained and no fuel damage will result. The
probability of fuel failure will not be increased by this proposed
TS change.
Therefore, the proposed TS change does not involve an increase
in the probability or consequences of an accident previously
evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed TS change will increase the High Pressure Coolant
Injection (HPCI) system response time from 30 seconds to 60 seconds.
This proposed change is bounded by the current Emergency Core
Cooling System (ECCS)--Loss-of-Coolant Accident (LOCA) analysis for
Limerick Generating Station (LGS) Units 1 and 2. The change in HPCI
system response time does not involve any physical modifications to
the plant systems or equipment, nor does it introduce a new
operational/failure mode, which might cause a different type of
accident. In case of a Loss of Feedwater Flow (LOFW) event, the HPCI
system will operate as designed, maintaining adequate core cooling.
Therefore, the proposed TS change does not create the
possibility of a new or different kind of accident, from any
accident previously evaluated.
3. The proposed TS change does not involve a significant
reduction in a margin of safety.
The following TS Bases were reviewed for potential reduction in
the margin of safety:
3/4.5 Emergency Core Cooling System
2.1.4 Reactor Vessel Water Level
The TS Bases do not discuss the High Pressure Coolant Injection
(HPCI) system start time. The margin of safety, as defined in the TS
Bases, will remain the same. The proposed TS change is in accordance
with the current licensing basis Emergency Core Cooling System
(ECCS)--Loss of Coolant Accident (LOCA) analysis for LGS Units 1 and
2, and does not impact any safety limits of the plant. The HPCI
system will operate as designed during the LOFW event, maintaining
adequate core cooling.
Therefore, the proposed TS change does not involve a reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101.
NRC Project Director: John F. Stolz.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
[[Page 35085]]
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of application for amendments: December 7, 1994.
Brief description of amendments: These amendments revise the Bases
of TS 3/4.7.5, ``Ultimate Heat Sink'' (UHS), to describe the UHS as
containing a 26-day supply of cooling water, instead of a 27-day
supply. In addition, the reference to Regulatory Guide 1.27 in the
bases of this TS would be revised to reference the January 1976
revision rather than the March 1974 revision.
Date of issuance: June 14, 1995.
Effective date: June 14, 1995.
Amendment Nos.: Unit 1--Amendment No. 93; Unit 2--Amendment No. 81;
Unit 3--Amendment No. 64.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the associated Bases of the Technical
Specifications.
Date of initial notice in Federal Register: March 1, 1995 (60 FR
11127) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 14, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of application for amendment: February 9, 1995.
Brief description of amendment: This amendment revises the reactor
high water level trip level setting for the Group 1 isolation. The
change will allow an increase to the main steam isolation valve high
water level isolation setpoint.
Date of issuance: June 15, 1995.
Effective date: As of the date of issuance to be implemented within
90 days.
Amendment No.: 164.
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 15, 1995 (60 FR
14017) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 15, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1
and 2, Will County, Illinois
Date of application for amendments: May 20, 1994, as revised on
February 2, 1995, and supplemented December 2, 1994, and March 14,
1995.
Brief description of amendments: The amendments revised the
Technical Specifications (TS) as they apply to Byron, Unit 1, and
Braidwood, Unit 1, to incorporate an alternative repair criteria for
defects found in the portion of the expanded steam generator tubes
within the tubesheet.
Date of issuance: June 22, 1995.
Effective date: June 22, 1995.
Amendment Nos.: 72, 72, 63, and 63.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: July 6, 1994 (59 FR
34659) and March 29, 1995 (60 FR 16184). The Commission's related
evaluation of the amendments is contained in a Safety Evaluation dated
June 22, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units
1 and 2, Rock Island County, Illinois
Date of application for amendments: September 15, 1992, as
supplemented April 21, 1995.
Brief description of amendments: This application upgrades the
current custom Technical Specifications (TS) for Dresden and Quad
Cities to the Standard Technical Specifications contained in NUREG-
0123, ``Standard Technical Specification General Electric Plants BWR/
4.'' This application upgrades only Sections 2.0 (Safety Limits and
Limiting Safety System Settings), 3/4.11 (Power Distribution Limits),
and 3/4.12 (Special Test Exceptions).
Date of issuance: June 13, 1995.
Effective date: Immediately, to be implemented no later than
December 31, 1995, for Dresden Station and June 30, 1996, for Quad
Cities Station.
Amendment Nos.: 134, 128, 155, and 151.
Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: May 10, 1995 (60 FR
24906) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 13, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: for Dresden, Morris Area
Public Library District, 604 Liberty Street, Morris, Illinois 60450;
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,
Illinois 61021.
[[Page 35086]]
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units
1 and 2, Rock Island County, Illinois
Date of application for amendments: December 15, 1993, as
supplemented April 21, 1995.
Brief description of amendments: These amendments upgrade the
current custom Technical Specifications (TS) for Dresden and Quad
Cities to the Standard Technical Specifications contained in NUREG-
0123, ``Standard Technical Specifications General Electric Plants BWR/
4.'' These amendments upgrade only Section 5.0 (Design Features). The
amendments include the relocation of some requirements from the TS to
licensee-controlled documents.
Date of issuance: June 14, 1995.
Effective date: Immediately, to be implemented no later than
December 31, 1995, for Dresden Station and June 30, 1996, for Quad
Cities Station.
Amendment Nos.: 135, 129, 156, and 152
Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: May 10, 1995 (60 FR
24909) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 14, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: for Dresden, Morris Area
Public Library District, 604 Liberty Street, Morris, Illinois 60450;
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,
Illinois 61021.
Consumers Power Company, Docket No. 50-255, Palisades Plant, Van Buren
County, Michigan
Date of application for amendment: December 13, 1994, as
supplemented May 3, 1995.
Brief description of amendment: This amendment revises the
Technical Specifications to add a high thermal performance (HTP)
departure from nucleate boiling correlation to Safety Limit 2.1. The
HTP correlation is used for HTP fuel loaded during recent fuel cycles.
Date of issuance: June 13, 1995.
Effective date: June 13, 1995.
Amendment No.: 168.
Facility Operating License No. DPR-20. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 10, 1995 (60 FR
24910) The May 3, 1995, submittal provided clarifying information which
was within the scope of the initial application and did not affect the
staff's initial proposed no significant hazards considerations
findings.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 13, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Van Wylen Library, Hope
College, Holland, Michigan 49423.
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver
Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania
Date of application for amendments: June 17, 1993, as supplemented
October 20, 1993, and May 23, 1995.
Brief description of amendments: These amendments revise the
Appendix A technical specifications (TSs) for Unit 1 and Unit 2 by
relocating the requirements of the radiological effluent technical
specifications (RETS) and the solid radioactive wastes TSs from the
Appendix A TSs to the offsite dose calculation manual (ODCM) or to the
process control program (PCP) in accordance with the guidance provided
in NRC Generic Letter 89-01 and NRC Report NUREG-1301. Programmatic
controls are also being incorporated into the Administrative Controls
section of the TSs. Additionally, editorial and definition changes are
being made to facilitate the relocation of these requirements.
Date of issuance: June 12, 1995.
Effective date: June 12, 1995.
Amendment Nos.: 188 and 70.
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 4, 1993 (58 FR
41504). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 12, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One Unit
No. 1, Pope County, Arkansas
Date of amendment request: May 15, 1995, as supplemented by letters
dated May 19 and June 7, 1995.
Brief description of amendment: The amendment was processed as an
exigent amendment following issuance of a notice of enforcement
discretion (NOED) by NRC letter dated May 17, 1995. The NOED and
exigent technical specification (TS) amendment authorized the licensee
to continue operating the reactor at power while the service water flow
to the reactor building emergency coolers is less than the TS
surveillance criteria.
Date of issuance: June 9, 1995.
Effective date: June 9, 1995.
Amendment No.: 182.
Facility Operating License No. DPR-51. Amendment revised the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: Yes (60 FR 27144, dated May 22, 1995). The notice
provided an opportunity to submit comments on the Commission's proposed
no significant hazards consideration determination. No comments have
been received. The notice also provided for an opportunity to request a
hearing by June 21, 1995, but stated that any such hearing would take
place after issuance of the amendment. The Commission's related
evaluation of the amendments, finding of exigent circumstances, and
final determination of no significant hazards consideration is
contained in a Safety Evaluation dated June 9, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: January 27, 1995.
Brief description of amendment: The amendment changed the Appendix
A Technical Specifications by increasing the allowable maximum
enrichment for the spent fuel pool and containment temporary storage
rack from 4.1 to 4.9 weight percent U-235 when fuel assemblies contain
fixed poisons.
Date of issuance: June 14, 1995.
Effective date: June 14, 1995.
Amendment No.: 108.
Facility Operating License No. NPF-38. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 15, 1995 (60 FR
14021)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 14, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.
[[Page 35087]]
Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie
Plant, Unit No. 2, St. Lucie County, Florida
Date of application for amendment: February 27, 1995.
Brief description of amendment: This amendment will modify
surveillance requirement (SR) 4.9.8.1 and 4.9.8.2 to allow a reduction
in the required minimum shutdown cooling flow rate under certain
conditions during operational MODE 6. In addition, the format of the SR
will be changed to clarify the intent of the stated surveillances.
Date of Issuance: June 14, 1995.
Effective Date: June 14, 1995.
Amendment No.: 76.
Facility Operating License No. NPF-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 29, 1995 (60 FR
16187) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 14, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of application for amendments: February 22, 1995.
Brief description of amendments: The proposed changes eliminate
reference to an automatic containment air lock tester from technical
specification 4.6.1.3. The automatic air lock tester is no longer being
used.
Date of Issuance: June 22, 1995.
Effective Date: June 22, 1995.
Amendment Nos.: 137 and 77.
Facility Operating License Nos. DPR-67 and NPF-16: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 29, 1995 (60 FR
16186) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 22, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant Units 3 and 4, Dade County, Florida
Date of application for amendments: January 17, 1995.
Brief description of amendments: These amendments concern
implementation of Florida Power and Light nuclear physics methodology
for calculations of the core operating limits report parameters.
Date of issuance: June 9, 1995.
Effective date: June 9, 1995.
Amendment Nos. 174 and 168.
Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 1, 1995 (60 FR
11133) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 9, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric
Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-424 and
50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,
Georgia
Date of application for amendments: October 3, 1994, as
supplemented by letter dated March 1, 1995.
Brief description of amendments: The amendments revise Technical
Specification 3/4.4.9, Pressure/Temperature Limits, and its associated
Bases, to provide new reactor coolant system heatup and cooldown
limitations and new power-operated relief valve setpoints for the low
temperature overpressure protection system.
Date of issuance: June 8, 1995.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 87 and 65.
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 21, 1994 (59
FR 65814) The March 1, 1995, letter provided supporting technical data
that did not change the scope of the October 1, 1994, application and
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated June 8, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Burke County Library, 412
Fourth Street, Waynesboro, Georgia 30830.
GPU Nuclear Corporation, Docket No. 50-320, Three Mile Island Nuclear
Station, Unit No. 2, (TMI-2), Dauphin County, Pennsylvania
Date of application for amendment: October 9, 1991.
Brief description of amendment: This amendment extends the
expiration date of the license from November 9, 2009 to April 19, 2014.
Date of issuance: June 21, 1995.
Effective date: June 21, 1995.
Amendment No.: 49.
Possession-Only License No. DPR-73: The amendment extends the
license expiration date.
Date of initial notice in Federal Register: August 3, 1994 (59 FR
39591). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 21, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, Walnut Street and Commonwealth
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
Gulf States Utilities Company, Cajun Electric Power Cooperative, and
Energy Operations, Inc., Docket No. 50-458, River Bend Station, Unit 1,
West Feliciana Parish, Louisiana
Date of amendment request: February 22, 1994, as supplemented May
19, 1995.
Brief description of amendment: The amendment revised Technical
Specifications 3.6.1.5, ``Main Steam--Positive Leakage Control
System,'' and 3.6.1.10, ``Penetration Valve Leakage Control System,''
to add an allowed outage time of 7 days with both trains of each system
inoperable. In addition, the allowed outage time for one train of the
Penetration Valve Leakage Control System inoperable is increased from 7
days to 10 days.
Date of issuance: June 19, 1995.
Effective date: June 19, 1995.
Amendment No.: 80.
Facility Operating License No. NPF-47. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 10, 1994 (59 FR
11331) The additional information contained in the supplemental letter
dated May 19, 1995, was clarifying in nature and thus, within the scope
of the initial notice and did not affect the staff's proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated June 19, 1995.
[[Page 35088]]
No significant hazards consideration comments received. No.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, Louisiana 70803.
PECO Energy Company, Public Service Electric and Gas Company, Delmarva
Power and Light Company, and Atlantic City Electric Company, Docket
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2
and 3, York County, Pennsylvania
Date of application for amendments: November 17, 1994 as
supplemented March 30, 1995.
Brief description of amendments: The amendments change equipment
designations, instrument range descriptions, instrument setpoints and
surveillance requirements in the Peach Bottom Technical Specifications
to reflect planned modifications to the main stack and vent stack
radiation monitoring systems.
Date of issuance: June 13, 1995.
Effective date: June 13, 1995.
Amendments Nos.: 204 and 207.
Facility Operating License Nos. DPR-44 and DPR-56: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 15, 1995 (60 FR
14027) The March 30, 1995, submittal provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 13, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
PECO Energy Company, Public Service Electric and Gas Company, Delmarva
Power and Light Company, and Atlantic City Electric Company, Docket
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2
and 3, York County, Pennsylvania
Date of application for amendments: March 16, 1995.
Brief description of amendments: These amendments change the
existing Technical Specification requirements for source range neutron
monitoring equipment while in the refueling mode to requirements based
on NUREG-1433, ``Standard Technical Specifications General Electric
Plants, BWR/4.''
Date of issuance: June 13, 1995.
Effective date: June 13, 1995.
Amendments Nos.: 205 and 208.
Facility Operating License Nos. DPR-44 and DPR-56: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 10, 1995 (60 FR
24913) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 13, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
PECO Energy Company, Public Service Electric and Gas Company, Delmarva
Power and Light Company, and Atlantic City Electric Company, Docket No.
50-277, Peach Bottom Atomic Power Station, Unit No. 2, York County,
Pennsylvania
Date of application for amendment: March 30, 1995, as supplemented
by letter dated May 26, 1995.
Brief description of amendment: The proposed amendment revises
Technical Specification Section 4.7.D.1.b(1) by adding a footnote to
exempt the High Pressure Coolant Injection motor-operated valve MO-2-
23-015 from quarterly stroke testing requirements until refueling
outage 2RO11.
Date of issuance: June 13, 1995.
Effective date: June 13, 1995.
Amendment No.: 206.
Facility Operating License No. DPR-44: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 10, 1995 (60 FR
24912) The May 26, 1995, submittal provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 13, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
PECO Energy Company, Public Service Electric and Gas Company, Delmarva
Power and Light Company, and Atlantic City Electric Company, Docket
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2
and 3, York County, Pennsylvania
Date of application for amendments: March 22, 1995.
Brief description of amendments: These amendments reduce the local
leak rate test hold time specified in the Technical Specification
Tables 3.7.2 through 3.7.4 from one hour to 20 minutes.
Date of issuance: June 19, 1995.
Effective date: June 19, 1995.
Amendments Nos.: 207 and 209.
Facility Operating License Nos. DPR-44 and DPR-56: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 10, 1995 (60 FR
24913). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 19, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of application for amendments: October 28, 1994, as
supplemented by letter dated April 18, 1995.
Brief description of amendments: These amendments delete, from the
Technical Specifications, the surveillance and operability requirements
for chlorine detection and the associated Bases as a result of the
removal of bulk quantities of gaseous chlorine from the site.
Date of issuance: June 19, 1995.
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment Nos.: 147 and 117.
Facility Operating License Nos. NPF-14 and NPF-22. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 21, 1994 (59
FR 65821). The April 18, 1995, letter provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination.
[[Page 35089]]
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 19, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701.
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: August 31, 1994.
Brief description of amendments: This amendment revises the
Technical Specifications to permit the operability requirement for the
Feedwater/Main Turbine Trip System Actuation Instrumentation to be
Operational Condition 1 greater than or equal to 25% Rated Thermal
Power.
Date of issuance: June 13, 1995.
Effective date: June 13, 1995.
Amendment Nos. 91 and 55.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 9, 1994 (59 FR
55884) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 13, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: August 23, 1994.
Brief description of amendments: Remove the 125/250 Vdc Class 1E
Battery Load Cycle Table from the technical specifications (TS) and
rephrase the surveillance requirements to be consistent with NUREG-
1433, ``Standard Technical Specifications'', and correct Amendments 71
and 34, dated June 28, 1994, to change certain surveillance requirement
intervals from 24 months to 18 months.
Date of issuance: June 19, 1995.
Effective date: June 19, 1995.
Amendment Nos. 92 and 56.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 12, 1994 (59 FR
51624) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 19, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: August 31, 1994.
Brief description of amendments: These amendments relocate the
requirements of TS 3/4.8.4.1, ``Primary Containment Penetration
Conductor Overcurrent Protective Devices,'' to the Updated Final Safety
Analysis Report and plant procedures.
Date of issuance: June 22, 1995.
Effective date: June 22, 1995.
Amendment Nos. 93 and 57.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 9, 1994 (59 FR
55884) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 22, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: August 12, 1994, as
supplemented by letter dated March 29, 1995.
Brief description of amendments: These amendments revise the action
statements regarding emergency core cooling systems to allow continued
operation in the event that the high pressure coolant injection system,
one core spray subsystem and/or one low pressure coolant injection
subsystem are inoperable.
Date of issuance: June 22, 1995.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos. 94 and 58.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 12, 1994 (59 FR
51623). The March 29, 1995, letter provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 22, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: August 31, 1994.
Brief description of amendments: The amendments permit the
operability of one Low Pressure Coolant Injection subsystem of Residual
Heat Removal while the subsystem is aligned and operating in the
Shutdown Cooling Mode during Operational Conditions (OPCONs) 4 and 5.
Date of issuance: June 22, 1995.
Effective date: June 22, 1995.
Amendment Nos. 95 and 59.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 9, 1994 (59 FR
55884). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 22, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New
Jersey
Date of application for amendments: November 18, 1994.
Brief description of amendments: The amendments revise the
Reactivity Control System Technical Specification Limiting Conditions
for Operation for boration flow paths and charging pumps by reducing
the number of operable charging pumps required for boron addition in
Mode 4 from two to one.
Date of issuance: June 12, 1995.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment Nos. 169 and 151.
Facility Operating License Nos. DPR-70 and DPR-75. The amendments
revised the Technical Specifications.
[[Page 35090]]
Date of initial notice in Federal Register: January 4, 1995 (60 FR
505). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 12, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, New Jersey 08079.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New
Jersey
Date of application for amendments: June 29, 1994, as supplemented
August 8, 1994, and May 2, 1995.
Brief description of amendments: The amendments increase the
Technical Specification minimum volume of emergency diesel generator
fuel oil contained in the Diesel Fuel Oil Storage Tanks at both units
of the Salem station.
Date of issuance: June 20, 1995.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment Nos. 170 and 152.
Facility Operating License Nos. DPR-70 and DPR-75. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 17, 1994 (59 FR
42346). The August 8, 1994, and May 2, 1995, letters provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 20, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, New Jersey 08079.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit
No. 1, Fairfield County, South Carolina
Date of application for amendment: March 6, 1995, as supplemented
on May 5, 1995 and June 6, 1995.
Brief description of amendment: The amendment deletes a license
condition that required the licensee to maintain a seismic monitoring
network around the Monticello Reservoir.
Date of issuance: June 13, 1995.
Effective date: June 13, 1995.
Amendment No.: 124.
Facility Operating License No. NPF-12. Amendment revises the
operating license.
Date of initial notice in Federal Register: March 29, 1995 (60 FR
16201). The May 5, 1995 and June 6, 1995 submittals provided
supplemental information that did not change the initial proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated June 13, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama
Date of application for amendments: November 15, 1994; superseded
March 7, 1995 (TS 350).
Brief Description of amendment: The amendments remove the
frequencies specified in the Technical Specifications for performing
audits and delete the requirement to perform the Radiological Emergency
Plan, Physical Security Plan, and Safeguard Contingency Plan reviews.
Date of issuance: June 19, 1995.
Effective Date: June 19, 1995.
Amendment Nos.: 221, 236 and 195.
Facility Operating License Nos. DPR-33, DRP-52 and DPR-68:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: December 21, 1994 (59
FR 65823); superseded March 29, 1995 (60 FR 16202). The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated June 19, 1995.
No significant hazards consideration comments received: None.
Local Public Document Room Location: Athens Public Library, South
Street, Athens, Alabama 35611.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: April 6, 1995 (TS 95-02).
Brief description of amendments: The amendments add a limiting
condition for operation that allows equipment to be returned to service
under administrative control to perform operability testing and
establishes the time interval to place an inoperable channel in the
bypass condition.
Date of issuance: June 13, 1995.
Effective date: June 13, 1995.
Amendment Nos.: 202 and 192.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: April 26, 1995 (60 FR
20530). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 13, 1995.
No significant hazards consideration comments received: None.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: April 6, 1995 (TS 95-05).
Brief description of amendments: The amendments revise the
technical specifications by deleting Tables 3.6-1, 3.6-2, and 3.8-2 and
referenced to them, incorporating related guidance and justification,
and modifying the specification related to electrical equipment
protective devices in accordance with Generic Letter 91-08.
Date of issuance: June 13, 1995.
Effective date: June 13, 1995.
Amendment Nos.: 203 and 193.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: May 10, 1995 (60 FR
24919). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 13, 1995.
No significant hazards consideration comments received: None.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: April 6, 1995 (TS 95-06).
Brief description of amendments: The amendments remove the
technical specification requirements related to crane travel over the
spent fuel pool.
Date of issuance: June 14, 1995.
Effective date: June 14, 1995.
Amendment Nos.: 204 and 194.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: April 26, 1995 (60 FR
20529). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 14, 1995.
[[Page 35091]]
No significant hazards consideration comments received: None.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of application for amendment: October 28, 1994.
Brief description of amendment: The amendment removes the Neutron
Monitoring System and Control Rod Position instrumentation from the
Vermont Yankee Technical Specifications for post-accident monitoring
and incorporates administrative changes.
Date of issuance: June 20, 1995.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 145.
Facility Operating License No. DPR-28. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 10, 1995 (60 FR
24922). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 20, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia.
Date of application for amendments: June 9, 1994.
Brief description of amendments: These amendments modify the
Chemical and Volume Control System and Safety Injection System
Technical Specifications.
Date of issuance: May 31, 1995.
Effective date: May 31, 1995.
Amendment Nos. 199 and 199.
Facility Operating License Nos. DPR-32 and DPR-37: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 20, 1994 (59 FR
37089). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 31. 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185.
Dated at Rockville, Maryland, this 27th day of June 1995.
For the Nuclear Regulatory Commission.
Jack W. Roe,
Director, Division of Reactor Projects--III/IV, Office of Nuclear
Reactor Regulation.
[FR Doc. 95-16249 Filed 7-3-95; 8:45 am]
BILLING CODE 7590-01-P